ML20212E691
| ML20212E691 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/21/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212E681 | List: |
| References | |
| NUDOCS 9909270114 | |
| Download: ML20212E691 (14) | |
Text
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UNITED STATES
-NUCLEAR REGULATORY COMMISSION g
g WASHINGTON, D.C. 20566-0001
,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON REVISED EMERGENCY ACTION LEVELS FOR FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302 -
1.0 INTRODUCTION
By letter dated July 29,1998, as supplemented by letters dated November 20,1998, July 13,1999,~ and August 31,1999, Florida Power Corporation (the licensee) submitted proposed changes to the Crystal River Unit 3 (CR-3) emergency action levels (EALs). The Office of Nuclear Reactor Regulation's review of these proposed changes against the regulatory requirements contained in Title 10, Code of Federal Reaulations (10 CFR) Part 50 is described below.
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2.0 BACKGROUND
The proposed revision to the CR-3 EALs was reviewed against the requirements in 10 CFR Section 50.47(b)(4) and Appendix E to 10 CFR Part 50.
It is specified in 10 CFR 50.47(b)(4) that onsite emergency plans must meet the following standard: "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee...."
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Appendix E, Subsection IV.B specifies in part that "These emergency action levels shall be discussed and agreed on by the applicant and State and local governmental authorities..."
Appendix E, Subsection IV.C specifies that ' emergency action levels (based not only on onsite and offsite radiation monitoring information but also on readings from a number of sensors that indicate a potential emergency, such as the pressure in containment and the response of the Emergency Core Cooling System) for notification of offsite agencies shall be described....
The emergency classes defined shall include: (1) notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency."
In Revision 3 to Regulatory Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors," the U.S. Nuclear Regulatory Commission (NRC) endorsed NUMARC/NESP-007, Revision 2, " Methodology for Development of Emergency Action Levels," as an acceptable method for licensees to meet the requirements of 10 CFR 50.47 (b)(4) and Appendix E to 10 CFR Part 50. The staff relied upon the guidance in NUMARC/NESP-007 as the basis for its review of the CR-3 EAL changes.
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. 3.0 EVALUATION The licensee's EALs are contained in a matrix which illustrates the progression of EALs from one classification level to the next. This matrix is supported by a basis document which includes, in addition to the EALs, the applicable operating mode, initiating conditions (ICs) and basis for each of the EALs. In addition, in its submittals, the licensee provided an attachment documenting deviations from the Nuclear Energy Institute (NEI) 97-03, Draft Revision 3
" Methodology for Development of Emergency Action Levels" EAL guidance.
The licensee stated that its revised EALs are based on NEl 97-03. NEl 97-03 is an industry-developed revision to the NUMARC/NESP-007 EAL guidance. As discussed previously, the NRC endorsed NUMARC/NESP-007 in Revision 3 of Regulatory Guide 1.101. The staff reviewed the licensee's EALs against the NUMARC/NESP-007 guidance. Any deviations from the NUMARC/NESP-007 guidance are discussed below. Where appropriate, the guidance provided in the proposed revision to NUMARC/NESP-007 in NEl 97-03 was considered in evaluating the adequacy of the CR-3 EALs.
The evaluations of the variations from the NUMARC/NESP-007 guidance are grouped into the
' following categories: (1) deviations from NUMARC/NESP-007, e.g., NUMARC/NESP-007 example EALs not included in the licensee's EAL scheme, (2) site-specific indications, e.g.,
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EALs which include unique thresholds or indications, and (3) site-specific additions, i.e., EALs not specified in the NUMARC/NESP-007 guidance which were included in the licensee's EAL scheme.
Deviation #1 AU2-3 Area Radiation Monitor Readings NUMARC/NESP-007 EAL AU2-3 reads as follows:
VALID Direct Area Radie!!on Monitortnadings increase by a factor of 1000 over normal levels.
The corresponding CR-3 EAL (EAL 1.7) reads as follows:
One or more valid radiation monitor readings unexpectedly exceed the values below for 15 minutes orlonger:
RM-G3 = 400 mR/hr RM-G4 = 600 mR/hr RM-G5 = 3000 mR/hr RM-G9 = 100 mR/hr RM-G10 = 800 mR/hr RM-G14 = 800 mR/hr RM-G17 = 800 mR/hr
3 The licensee's EAL deviates from the NUMARC/NESP-007 guidance by incorporating the condition of "exceading the values below for 15 minutes." Incorporating the 15-minute condition in this EAL will discriminate against transient conditions where dose rates are immediately reduced below the setpoint. These transient conditions do not meet the definition of an Unusual Event. Therefore, this deviation is acceptable. The CR-3 EAL corresponding to NUMARC/NESP-007 EAL AA3-2 deviates in a similar manner and is also acceptable.
Deviation #2 AG1-3 Dose Assessment NUMARC/NESP-007 AG1-3 reads as follows:
. A valid dose assessment capabilityindicates dose consequences greater than 1000 mR whole body or5000 mR child thyrord The corresponding CR-3 EAL (EAL1.4-2) reads as follows:
Dose assessment results indicate site boundary doses greater than 1000 mR TEDE
[ total effective dose equivalent] or 5000 mR thyroid CDE [ committed dose equivalent]
forthe actual orprojected duration of the release and core damage is suspected orhas l.
occuned.
l The licensee's EAL deviates from the NUMARC/NESP-007 guidance by incorporating the j
cond'. tion that " core damage is suspected or has occurred." The licensee stated that it -
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' incorporated this condition because " verifying core damage... precludes erroneous protectiw action recommendations based on incorrect or default dose assessments when plant I
conditions clearly do not support the magnitude of the release." Furthermore, the licensee l
stated that " based on FSAR [ Final Safety Analysis Report) isotopic distribution for accidents and very conservative ' worst case' meteorological data, PAG [ protective action guideline] limits j
cannot be reached without some amount of fuel damage."
The staff agrees that classification of events should be as accurate as possible, espscially at l
the higher classification levels. Therefore, the staff concludes that including this check for core l
~ damage so as to ensure accurate classification is acceptable.
Deviation #3 Site Area Emergency Classification - Fission Product Barrier Combination The NUMARC/NESP-007 guidance provides a fission product matrix for classifying events based upon the loss or potentialloss of combinations of fission product barriers. The site area emergency is classified based upon the following combinations of barriers.
Loss of BOTH Fuel Clad and RCS [ reactor coolant system]
OR j
Potential Loss of BOTH Fuel Clad and RCS i
OR Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Banier J
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. The CR-3 EAL scheme classifies a Site Area Emergency whenever any two barriers are lost or potentially lost. This is consistent with the revised EAL guidance provided in NEl 97-03. The
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- method for classifying the Site Area Emergency was changed in order to simplify the logic. A review of the potential combinations of EALs indicates that the indications (i.e., thresholds) for j
the potential loss of the containment barrier provided in the CR-3 EAL scheme (e.g., reactor j
building pressure > 54 psig, hydrogen concentration, and containment radiation levels) correspond to conditions of the loss of fuel clad or RCS. Therefore, existence of these conditions warrants classification of a Site Area Emergency under the NUMARC/NESP-007 EAL guidance. The CR-3 site-specific method for classifying a Site Area Emergency will result in the proper classification of events based upon the combination of fission product barriers impacted. Therefore, the staff considers this deviation acceptable.
l Deviation #4 Fission Product Barrier-Critical Safety Function Status The NUMARC/NESP-007 guidance provides example EALs based upon Critical Safety i
Function Status (CSFS) indication. CSFS, which are used at Westinghouse nuclear power plants, are not used at CR-3 (a Babcock and Wilcox Nuclear Power Plant). Therefore, the CR-3 EAL scheme does not include these NUMARC/NESP-007 example EALs. The parameters which provide input to the Westinghouse CSFS are utilized as EALs in the CR-3 EAL scheme. Therefore, this deviation is acceptable.
Deviation #5 Fission Product Barrier-Core Exit Thermocouple Temperature One of the NUMARC/NESP-007 EALs for the loss of the fuel clad barrier reads as follows:
GREATER THAN (site-specific) degree F The corresponding CR-3 EAL reads as follows:
Core Conditions in Region 3 or severe accident mgion ofICC (inadequate Core
-Cooling] curves The intent of the NUMARC guidance is to specify an EAL which is an indi:ation of clad damage.
The CR-3 ICC cunres define regions of reactor pressure and core exit thermocouple temperature corresponding to certain fuel clad temperatures. Region 3 corresponds to reactor pressure and temperature conditions where the clad temperature exceeds 1400 *F. This clad temperature corresponds to possible gap release. The severe accident region corresponds to reactor pressure and temperature conditions where the clad temperature exceeds 1800 *F.
This clad temperature corresponds to possible core melt. The use of the ICC curves in place of a rirc;le core exit thermocouple reading (as specified in NUMARC) meets the intent of the corresponding NUMARC/NESP-007 EAL and, therefore, is acceptable.
Deviation #6 Fission Product Barrier Matrix - Potential Loss of Fuel Clad based upon Vessel Water Level
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.- One of the NUMARC/NESP-007 EALs for the loss of the fuel clad barrier based upon reactor vessel waterlevel reads as follows:
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Level LESS than (site-speci6c) value The licensee included the following EAL in place of this NUMARC/NESP-007 EAL:
RCS conditions wanant entryinto EOP [ Emergency Operating Procedure]-07 l
The licensee's rationale for this deviation is:
Entryinto EOP-07, ' Inadequate Core Cooling"is a clearindicator that there are i
superheated conditions in the core which maylead to clad damage. Superheated conditions indicate that the waterlevelis below the top of active fuel. Entry conditions forEOP-07is 20 *F superheat.
The NUMARC guidance specifies that the site-specific value should correspond to indication of a challenge to core cooling. The licensee's EAL is consistent with the intent of the NUMARC/NESP-007 guidance and, therefore, is acceptable.
1 Deviation #7 Fission Product Barrier Matrix - Loss of RCS Earrier based upon Indication of l
Steam Generator Tube Rupture The NUMARC/NESP-007 EAL for the loss of the RCS barrier based upon Indication of Steam Generator Tube Rupture is:
(Site-speci6c) indication that a SG [ steam generator]is mptured and has a non-isolable
. secondaryline break <0R> (site-speci6c) indication that a SG is ruptured and a prolonged release of contaminated secondary coolant is occurring from the affected SG to the environment The licensee included the following EAL in place of this NUMARC/NESP-007 EAL:
OTSG [once-through steam generator] tube leak results in loss of adequate subcooling l
margin I
The CR-3 EAL deviates from NUMARC in that it does not include a statement regarding a
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release to the environment. The CR-3 EAL is consistent with a revision to this EAL guidance l
provided in NEl 97-03. The change in NEl 97-03 better correlated the EAL to the condition of concem, i.e., loss of the RCS barrier. The CR-3 EAL provides a valid indication of the condition
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l of concem, i.e., loss of RCS barrier. Therefore, this deviation is acceptable.
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Deviation #8 Fission Product Barrier-Loss of Containment based upon Indication of Steam
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Generator Tube Rupture The NUMARC/NESP-007 EAL for the loss of the containment barrier based upon Indication of Steam Generator Tube Rupture reads as follows:
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pi Release of secondary side to atmosphere with primary to secondaryleakage
' GREATER THAN tech spec allowable
- The licensee included the following EAL in place of this NUMARC/NESP-007 EAL:
OTSG has Tube Leak per EOP-06, SGTR (steam generator tube rupture) and unisolable steam leak outside RB Operators enter EOP-06 upon a steam generator tube leak of 1 gpm. NEl 97-03 revised the NUMARC guidance for this EAL and raised the leak rate to 10 gpm because leakage at technical specification limit levels was considered to be too low to constitute a loss of the containment barrier. The staff considers the use of a 1-gpm threshold for this foss of containment EAL to be appropriate. CR-3 EAL provides an appropriate indicator of the condition of concem, i.e., loss of containment barrier. Therefore, this deviation is acceptable.
Deviation #9 Fission Product Barrier-Potential Loss of Containment Barrier based upon
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Core Exit Thermocouple Readings The NUMARC/NESP-007 EAL for the potentialloss of the containment barrier based upon Core Exit thermocouple readings is:
l Ct.,'e exit thermocouples in excess of 1200 degrees and restoration procedures not effectin within 15 minutes; or, core exit thermocouples in excess of 700 degrees with reactor vessellevel below top of active fuel and restoration procedures not effective
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within 15 minutes The corresponding CR-3 EAL is:
l Core conditions in severe accident region ofICC curves for > 15 minutes As discussed under Deviation #5, the severe accident region of the ICC curves corresponds to clad temperatures exceeding 1800 'F which indicate that a core melt condition exists. This is consistent with the intent of the NUMARC/NESP-007 EAL and, therefore, is acceptable.
Deviation #10 HU4 - Relocation of Security Related EAL (Bomb)
NUMARC IC HU4 reads as follows:
Con 6tmed security event which indicates a potential degradation in the level of safety l
of the plant One of the NUMARC/NESP-007 example EALs under this IC reads as follows:
Bomb device discovered within plant Protected Area and outside the plant Vital Area I
L l' l The CR-3 EAL scheme did not include this under its Unusual Event Security IC but did include a similar EAL as an Alert (EAL 2.18). A bomb discovered within the protected area meets the l
condition specified in the NUMARC Alert level Security IC HA4, i.e., " Security Event in the Protected Area.". Therefore, this deviation is acceptable.
Deviation #11 : HA2 - Fire NUMARC IC HA2 reads as follows:
Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.
One of the EALs under this IC roads as follows:
Fire or explosion in any of the following (site-specific) areas:
(site-specific) list AND Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equip?ent within the specified area The corresponding CR-3 IC for the fire portion of the NUMARC IC reads as follows (note that separate ICs were provided for a fire and an explosion):
FIRE athcting the Operability of SAFE SHUTDOWN EQUIPMENT The corresponding CR EAL reads as follows:
Report by plant personnel of VISIBLE DAMAGE to SAFE SHUTDOWN EQUIPMENT due to FIRE OR Indications show degraded SAFE SHUTDOWN EQUIPMENTperformance due to the FIRE The basis for the CR-3 includes the following statement which clarifies the EAL.
If damage from the fire is clearly contained and localized to one train and safe shutdown capability exists, then the EAL is not met. If the extent of the damage is uncertain in terms ofloss of sak shutdown capability, then entryinto this EAL is required.
This statement restricts this EAL so that it is only applicability to fires which have the potential to propagate or affect the ability to safely shut down the plant. This is appropriate because if the fire is contained and does not affect the ability of the plant to shut down it does not represent a substantial decrease in the safety of the plant (and therefore does not warrant an Alert classification). The additional condition provided in the basis to this EAL assists the emergency director in accurately classifying a fire event. Therefore, this deviation is acceptable. A similar statement is included in the basis for the vehicle crash, explosion, and
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- intemal flooding EALs, These EALs are similar in nature, i.e., the concem is degradation of the ability to safely shut down the plant and the effect of the event may or may not be j
significant. This statement will assist in classifying the severs events as Alerts while preventing over-classifying events which are not a substantial degradation in the level of safety of the plant. Therefore, including this statement in the basis is a!so appropriate for these EALs.
i Deviation #12 HA3 - Toxic Gas Levels l
NUMARC IC HA3 reads as follows:
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Release of Toxic orFlammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown One of the EALs under this IC reads as follows:
Report or detection of toxic gases within a facility structure in concentrations that will be life threatening to plant personnel.
l The corresponding CR-3 EAL reads as follows:
Toxic Gas levels > IDLH [immediately dangerous to life orhealth] levels in areas that I
require continuous occupancy to maintain sak operation or establish ormaintain cold shutdown 1
OR Toxic Gas lents >IDLH levels within the Protected Area such that plant personnel are unable to perform actions necessary to maintain sak operation or establish or maintain cold shutdown using protective equipment.
The CR-3 EAL scheme deviates from the NUMARC guidance by including one EAL which includes consideration of the use of protective equipment in determining whether the IC is met.
This EAL is only applicable in those areas that do not require continuous occupancy. The EAL is an effective threshold for the condition of concem identified in the IC and therefore this deviation is acceptable.
Deviation #13 -
HG1 - Security Event NUMARC/NESP-007 IC HG1 is:
1 Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown.
The corresponding CR-3 IC reads as follows:
Security Event Resulting in Loss of Physical Control of the Facility.
The NUMARC/NESP-007 EALs under IC HG1 read as follows-j i
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. Loss of physical control of the control room due to security event.
Loss of physical contml of the remote shutdown capability due to secun'ty event.
The corresponding CR-3 EAL (2.21) reads as follows:
Intruder (s) has taken control of the ControlRoom, or Remote Shutdown Room orplant equipment such that plant personnel are unable to operate equipment required to establish and maintain safe shutdown conditions.
The IC was modified to more-closely reflect the plant condition of concem (as reflected in the EALs under this IC). The EAL was modified to recognize that a loss of physical control of the plant varies based upon plant specific configurations of equipment and controls. The basis section of this EAL was modified to provide additional guidance on the site-specific implementation of this EAL. The intent of the IC and EAL was not changed. Therefore, the IC and EAL are appropriate.
Deviation #14 SU4
- Fuel Clad Degradation NUMARC lC SU4 reads as follows:
Fuel Clad Degradation The NUMARC/NESP-007 EALs under this IC read as follows:
(Site-speci6c) radiation monitor readings indicating fuel clad degradation greater than technicalspeci6 cation limits.
OR (Site-speci6c) coolant sample activity value indicating fuel clad degradation greater than technicalspeci& cation limits.
The CR-3 did not include an EAL which corresponds to the first EAL under this IC. CR-3 did not include this EAL because it does not have a failed fuel monitor. The CR-3 plant has a liquid RCS monitor; however, this radiation monitor does not provide an indication of fuel degradation greater than technical specification limits. The licensee states that an alarm on j
this monitor would trigger an RCS sample to be taken for use in classifying the event (the CR-3 EAL scheme does include an EAL corresponding to the second EAL under this IC).
Furthermore, RCS samples are taken on a daily basis. Because the licensee does not have a l
radiation monitor to use to develop an EAL corresponding to the NUMARC guidance, this i
deviation is acceptable.
Site-Soecific Indication #1 AS1-1 Radiation Monitor Setpoint NUMARC/NESP-007 EAL AS1-1 reads as follows:
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. 1 A valid reading on one or more of the following monitors that exceeds oris expected to exceed the value shown indicates that the release may have exceeded the above
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criterion andindicates the need to assess the release with (site-specific procedure):
1 (site-specinclist)
The CR-3 EAL corresponding to this EAL is:
l Valid RM-A1 orRM-A2 Mid-Range monitor reading exceeds the values on the following table for the current stability class for 15 minutes orlonger Stab. Class Reading (mR/hr)
A, B, or C 80 D ore 20 F or G 5
I The CR 3 EAL provides three setpoints based upon the current stability class. This provides for a closer correlation of the radiation monitor setpoint to the IC which is based upon predicted dose consequences. These setpoints are consistent with the NUMARC/NESP-007 guidance and are acceptable.
Site-Soecific Indication #2 Potential Loss of RCS Barrier based upon Emergency System Actuation The licensee added the following EALs as indicators of the potentialloss of the RCS barrier:
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RCS leak results in ES [ engineered safeguards] Actuation on Low RCS Pressure OTSG Tube leak results in ES Actuation on Low RCS Pressure The licensee's rationale for including these EALs is:
Should the injection system fail or the operator fail to open the injection valves upon a failure of the Makeup system to mair:tain RCS inventory, RCS pressure will decrease ts the ES actuation setpoint. This potential 10ss factorin addition to number one (abon) will ensure the RCS barrier will be considered potentially lost for any inability of the makeup system to maintain adequate inventory during a loss of coolant event.
These site-specific EALs are indicators of the potentialloss of the RCS barrier and, therefore, are acceptable.
Site-Specific Indication #3 SS5 - Loss of Vessel Water Level NUMARC IC SS5 reads as follows:
Loss of Water Levelin the Reactor Vessel That Has or Will Uncover Fuelin the Reactor Vessel.
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. The NUMARC/NESP-007 EAL under IC SSS reads as follows:
Loss of all decay heat removal cooling as determined by (site-specific) procedure.
AND i
(Site-specific) indicators that the core is or will be uncovered.
The corresponding CR-3 EAL reads as follows:
1.
Loss of decay heat removalper AP-404 AND 2.a incores indicating superheated conditions OR 2.b incores unavailable and time to uncovery exceeded as specified in OP-301 The CR-3 EAL did not use vessel level indication but rather used a site-specific indication for core uncovery. The licensee did not use vessellevel indication because its vessellevel indication system only provided a measurement of water level in the reactor vessel head region. The licensee utilized two conditions in place of vessellevel indication. The first condition indicates that reactor water levelis below top of active fuel. The second condition is useful to predict whether reactor vessel water level is below the top of active fuel when the incores are not available. The CR-3 site-specific indication provides an effective threshold for the IC. Therefore, this site-specific indication is acceptable.
Site Specific Addition #1 Liquid Effluent Release The licensee added the following EAL under its IC for unplanned release of liquid radioactivity to the environment (which corresponds to NUMARC/NESP-007 IC AU1):
Release continued for 60 minutes orlonger with no dilution flow This added EAL is a valid indication of the plant IC of concem, i.e., the loss of control of radioactive materials. Therefore, this EAL is acceptable.
Site-Soecific Addition #2 Hurricane Waming The CR-3 EAL scheme includes the following site-specific EAL classified at the UE level under the category of Natural / Man-Made Hazards:
Hum' cane Waming This EAL is included in the m rer>tly approved EAL scheme (Revision 18) and initiates
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emergency response actions in accordance with the CR-3 emergency plan in preparation for the hurricane. A hurricane may be a precursor to more severe events (e.g., loss of offsite power). This condition is consistent with the definition of an UE. Therefore, this site-specific EAL is acceptable.
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l Site-Soecific Addition #3 Flooding The CR-3 EAL includes the following site-specific EAL under the category of Natural / Man-t l
Made Hazards:
l Intake canallevel or visual observation indicates flood waterlevel 198 feet l-This condition, classified at the UE level, is considered to be a precursor of more serious
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events. At 98 feet water level, there is no immediate impact on plant equipment but heightened awareness is appropriate. This condition is consistent with the definition of an UE.
l Therefore, this site-specific EAL is acceptable.
Site-Soecific Addition #4 Intemal Flooding The CR-3 EAL includes the following site-specific EALs under the category of Natural / Man-Made Hazards:
Unusual Event Indication of uncontrolled flooding in the Auxiliary Building orIntermediate Building AND WaterlevelMiooding has the potential to affect orimmerse Safe Shutdown Equipment Alert
- 1. Waterlent exceeds 1.5 feet in the auxiliary building orintermediate building AND 2.a Indications show degraded safe shutdown equipment due to the flooding OR 1
2.b Electrical hazards prennt plant personnel normal access to areas of plant j
containing safe shutdown equipment The plant condition specified in the Unusual Event EAL is indication of the potential J
degradation in the level of safety of the plant. In addition, the plant condition in the Alert EAL is indication of a substantial degradation in the level of safety of the plant. These EALs are consistent with the definitions of an UE and Alert. Therefore, these site-specific EALs are acceptable.
Site-Scecific Addition #5 Inadvertent Positive Startup Rate The CR-3 EAL includes the following site-specific EAL under the category of System Malfunctions:
An extended or UNPLANNED positive starfup rate monitored by nuclear anstrumentation The plant condition specified in the UE EAL is indication of the potential degradation in the level of safety of the plant in that reactor reactivity is not properly controlled. This condition is consistent with the' definition of an UE. Therefore, this EAL is acceptable.
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4.0 CONCLUSION
The proposed EAL changes for CR-3 are consistent with the guidance in NUMARC/NESP-007, with variations as identified and accepted in this review, and, therefore meet the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR Part 50.
Principal Contributors: J. O'Brien L Cohen 1
Date: September 21, 1999 i
i Mr. John Paul Cowan CRYSTAL RIVER UNIT NO. 3 Florida Power Corporation cc:
Mr. R. Alexander Glenn Chairman Corporate Counsel (MAC-BT15A)
Board of County Commissioners J
Florida Power Corporation Citrus County I
P.O. Box 14042 110 North Apopka Avenue St. Petersburg, Florida 33733-4042 invemess, Florida 34450-4245 Mr. Charles G. Pardee, Director Ms. Sherry L. Bernhoft, Director Nuclear Plant Operations (PA4A)
Nuclear Regulatory Affairs (NA2H)
Flonda Power Corporation.
Florida Power Corporation Crystal River Energy Complex Crystal River Energy Complex l
15760 W. Power Line Street 15760 W. Power Line Street Crystal River, Florida 34428-6708 Crystal River, Florida 34428-6708 Mr. Michael A. Schoppman Senior Resident inspector Framatome Technologies Inc.
Crystal River Unit 3 1700 Rockville Pike, Suite 525 U.S. Nuclear Regulatory Commission 4
Rockville, Maryland 20852 6745 N. Tallahassee Road Crystal River, Florida 34428-Mr. William A. Passetti, Chief j
Department of Health Mr. Gregory H. Halnon Bureau of Radiation Control Director, Quality Programs (SA2C) 2020 Capital Circle, SE, Bin #C21 Florida Power Corporation
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Tallahassee, Florida 32399-1741 Crystal River Energy Complex 15760 W. Power Line Street Attorney General Crystal River, Florida 34428-6708 Department of Legal Affairs j
The Capitol j
Tallahassee, Florida 32304 l
Mr. Joe Myers, Director
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Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 l
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