IR 05000293/1998007
| ML20154J905 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 10/07/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154J893 | List: |
| References | |
| 50-293-98-07, 50-293-98-7, NUDOCS 9810160034 | |
| Download: ML20154J905 (29) | |
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Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION l '
License No.: OPR-35 Report No.: 98-07 Docket No.: 50-293 Licensee: BEC Energy 800 Boylston Street Boston, Massachusetts 02199 Facility: - Pilgrim Nuclear Power Station Inspection Period: July 26,1998 - September 6,1998 Inspectors: R. Laura, Senior Resident inspector R. Arrighi, Resident inspector ,
S. Dennis, DRS Inspector K. Kolaczyk, DRS Inspector D. Silk, Regional Specialist ,
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Approved by: C. Cowgill, Chief' ,
Reactor Projects Branch No. 5 Division of Reactor Projects
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9810160034 981007 PDR
G ADOCK 05000293 PDR -
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EXECUTIVE SUMMARY Pilgrim Nuclear Power Station NRC Inspection Report 50-293/98-07 1
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i This integrated inspection included aspects of licensee operations engineering , ,
maintenance, and plant support. The report covers resident inspection for the period of i July 26,1998, through September 6,1998;in addition, it includes the results of l announced inspections by DRS inspectors. In addition, it includes an in-office review of ;
- licensee emergency procedure changes. I i
Operations '
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.* Routine plant operations were performed well, especially during several
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l downpowers to backwash the condenser water boxes. An increased use of 1
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industry operating experience was evident at the Plant Manager's morning meeting. !
A cross functional self assessment utilized industry peers and was self criticalin ;
identifying areas for improvement. (Section 01.1) !
- - The inspectcM noted three occasions when degraded equipment conditions were j either not expeditiously identified or communicated. Plant and operations e management acknowledged the observation and stated there was an ongoing effort l
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to strengthen management expectations in this area. (Section 02.1) l
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- Operations personnel performed fuel movements well in the spent fuel pool, with
- appropriate concerns for safety, and foreign materials exclusion. Communications between operator, spotter, and reactor engineering personnel were clear and concise. (Section 04.1)
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Maintenance ,
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- - Use of mock-up training and extra electrical work precautions contributed to j successful replacement of the electrical brushes on the exciter for the "B" motor i
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generator set motor. (Section M1.1) i tk A
- The maintenance staff exhibited conservative decision making by placing the work to replace the "B" control rod drive (CRD) pump casing gasket on hold in lieu of ,
attempting to perform the work with boundary valve leakage. (Section M1.1)
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i. * Surveillance testing was done in a well controlled manner consistent with safety
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requirements. Testing was sufficient in scope to demonstrate that the subject equipment would perform their safety 'unmions. Engineering and radiation
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protection personnel provided the nneesary support of testing activities. (Section M1.1)
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e' The inspector determined that the performance and scheduling of selected
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inftequently performed technical specifications required surveillances were
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accurately tracked in the Master Surveillance Tracking Program for the core and
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containment cooling,125VDC, 250VDC, core spray, standby gas treatment, and drywell/ torus header systems. (Section M1.2)
- The inspector determined that the temporary repair of (salt service water) SSW spool piece JF29-8-4 was adequately performed and that the operability determination was appropriate in scope and detail. A thorough review of the work package by quality assurance personnelidentified that the incorrect weld metal was used in the repair of the SSW pipe. (Section M2.1)
- The engineering staff properly utilized NRC GL 90-05 to develop and execate the non-ASME (American Society of Mechanical Engineers) repair plan for the degraded SSW (salt service water) spool piece. The inspection of nine similar pipe locations for erosion / corrosion was very rigorous. (Section M2.1)
- Improved material condition was noted in the reactor building control rod drive quadrant room due to increased management attention. Also, new permanent lighting was added to the torus room. (Section M2.2) l Enaineerina
- The NPO and the licensee's engineering staff identified deficiencies in the design of the control room high efficiency air filtration system, i.e., seismic qualification of duct supports, single failure design analyses, and inconsistencies between analysis assumptions and operational procedures. The deficiencies have been reported in two licensee event reports, including acceptable short term corrective actions. The failure to accurately transfer the system design into instructions, procedures or drawings was a violation of 10 CFR 50 Appendix B Criterion lli " Design Control."
(Section E2.1)
- An NRR safety evaluation agreed with the licensee resolution of non-conforming jet pump swing gate and restraint issues.
Plant Suocort
- The circumstances leading to the misplacement of the intermediate range detectors represented a lack of accountability and control of special nuclear material. The licensee was successfulin finding the missing intermediate range monitors, however the search resulted in an accumulated dose of approximately one rem.
The corrective actions to resolve the issue were determined to be appropriate.
(Section S1.1)
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TABLE OF CONTENTS EX EC UTI VE S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii Summary of Plant Status ............................................1 l l 1. O P E R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 01' Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comments (71707) ...........................1 !
02 Operational Status of Facilities and Equipment ................... 2 l O2.1 Operator Awareness and Communication of Equipment Status . . . 2 l 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 3 04.1 Observation of Spent Fuel Pool Operations . . . . . . . . . . . . . . . . . 3 l
08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
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08.1 (Closed) LER 5 0-2 9 3 /9 7-04-01 . . . . . . . . . . . . . . . . . . . . . . . . . 3 08.2 (Closed) LER 5 0-2 9 3 /9 8-0 6-01 . . . . . . . . . . . . . . . . . . . . . . . . . 4 11. M A I NT E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M 1.1 General Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . 4 l M1.2 Surveillance Frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 7 l M 2.1 Non-ASME Code Repair on Salt Service Water (SSW) Piping . . . . . 7 M2.2 (Closed) VIO 98-02-03: Degraded Equipment Problem identification
...............................................8 111. EN G I N E ERI N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 8 E2 Engineering Support of Facilities and Equipment .................. 8 E2.1 (Clos.ed) LERs 50-293/98-08 and 98-16 (Open) VIO 50-293/98-02 . 8 E2.2 (Closed) VIO 98-02-07: Inadequate Design Control . . . . . . . . . . 11 l
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E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 E8.1 (Closed) IFl 97-02-03: Jet Pump Nonconforming Conditions . . . . 12 I V . P LA N T S U PP O RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 1 P3 EP Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . 12 S1.1 Control of Special Nuclear Material .....................12 V. M AN AG EM ENT M EETIN G S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
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INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 ITEMS OPENED, CLOSED, AND UPDATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 i
LI ST O F AC RO NYM S U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
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i ATTA C H M E NT 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 8 ATTA C H M E NT 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , , , , , , , , , , , , , , , , , 2 3 i
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REPORT DETAILS !
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Summarv of Plant StJL19_f I Pilgrim Nuclear Power Station (PNPS) began the period at.50 percent power for a planned u thermal backwash of the main condenser to reduce biofouling. Full power was attained on
! July 26,1998. Several times throughout the period, power was again lowered to perform ,
' a backwash due to the buildup of biofouling in the condenser water box. The plant was at t
'100 percent power at the end of'the report period.
1. OPERATIONS-01 Conduct of Operations'
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. 01.1 General Comments (71707)
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Plant operations were conducted with a proper focus on nuclear safety. The
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inspector verified during daily control room observations that operators remained attentive and responsive to plant conditions. Daily plant operations were conducted i
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. reviews of ongoing plant operations and observed that overall plant staff conduct -
j. was professional and safety-conscious. Operators were aware of the status of
. plant equipment, properly used three-way communications, and effectively implemented plant procedures, i
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l Operators were required to monitor biofouling in the main condenser by closely
! trending the ' condenser hotwell temperature to ensure the temperature remained j l . below the 120 degree administrative limit and above 26 inches of vacuum in the '
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condenser. On several occasions operators lowered reactor power to stay below ,
the limit to prevent condensate pump cavitation. Also, operators conducted several backwash operations of the condenser waterboxes which required lowering reactor !
!. power to approximately 50%.. The inspector determi.W that operators performed i very well during the power changes to cope with the effects of condenser i biofouling. I L ~ A review of the standby line-up of the standby gas treatment (SBGT) was L performed including electrical breakers, dampers and indications. The system was j
t l ' found to be properly lined up to support operability.
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'The Plant Manager's morning meeting included a new initiative to review recent operating experience information. This was implemented by having one manager review the current operating experience reports and select one event to be briefed at the morning meeting. The inspector attended several briefings and noted the
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- events were covered in detail and specifically included a review for applicability at .
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!- ! Topical headings such as 01, MS, etc., are used in accordance with the NRC ;
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standardized reactor inspection report outline. Individual reports are not expected to address all outline topics.
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Pilgrim. In addition, A one-week licensee self assessment was conducted which utilized industry peer members to conduct a cross functional assessment.
I The fire brigade performed well during an unannounced fire drillin the reactor l l building. The operators assigned to the brigade arrived at the scene donned in fire protection equipment within approximately 10 minutes. The drillinstructors provided close oversight of the operator response. l In summary, routine plant operations were performed well especially during several I downpowers to backwash the condenser water boxes. An increased use of industry operating experience was evident at the Plant Manager's morning meeting.
A cross functional self assessment utilized industry peers and was self criticalin ,
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identifying areas for improvement. l 02 Operational Status of Facilities and Equipment O2.1 Operator Awareness and Communication of Eauioment Status Information a. Insocction Scope .
The inspector verified that actual equipment conditions were accurately dUcussed during shift briefings and documented in the daily operating report.
b. Observations and Findinas Out of service equipment was properly tracked in the terhnical specification logbook maintained in the control room. Degraded equipment was typically discussed at shift briefings and captured in the daily operating report. The inspector noted'three inconsistencies with the identification and/or con.nunication of equipment status information.
An operations daily morning report listed mechanical sealleakage from a reactor water clean-up system pump as 45 drops per minute when the leakage had degraded to approximately one gallon per minute. The inspector determined that the degraded seal leakage was not properly documented in the daily report or discussed at the daily management morning meeting. Also, a shift briefing did not include relevant information regarding the status of a control rod drive pump maintenance activity.
l l The third event occurred on August 15 when the turbine building tour operator l observed that the auxiliary oil pump for the "C" reactor feed pump was unexpectedly operating. Normally, the auxiliary lubricating oil pump remains i secured in the standby condition. Operators secured the auxiliary oil pump; however, the pump automatically restarted. After further investigation, the licensee identified a broken sensing line that caused the purnp to start. The licensee implemented a temporary modification until the broken sensing line could be repaired. The inspector noted that the field operator identification of this condition demonstrated good attention to detail. The inspector noted that this condition had (
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c. Conclusions
! The inspectors noted three occasions when degraded equipment conditions were either not expeditiously identified or communicated. Plant and operations management acknowledged the observation and stated there was an ongoing effort to strengthen management expectations in this area.
04- Operator Knowledge and Performance
.04.1 Observation of Soent Fuel Pool Operations Insoection Scoos (71707)
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The inspector observed and assessed spent fuel movements in the spent fuel pool i L (SFP) during the search for missing intermediate range meters (IRMs).-
a b; Observations and Findinas
--The inspector observed the movement of three fuel bundles that were moved during
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the search for the four missing IRMs (refer to Section S1.1). The crane operator was observed to be experienced and knowledgeable in the operation of the SFP bridge crane. Foreign material exclusion practices were used by all personnel
- entering the controlled area around the SFP. Repeat back communication was observed between the operator, spotter, and the reactor fuels engineer responsible -
for inventory controls of the spent fuel. The inspector verified that the licensee followed the applicable plant procedures and that technical specification
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c. Conclusions
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Operations personnel performed fuel movement well, with appropriate concerns for safety, and foreign materia's exclusion. -Communications between operator, spotter, and reactor engineering personnel were clear and concise.
08 Miscellaneous Operations issues (92700,92901)
- 08.1- (Closed) LER 50-293/97-04-Q1: Loss of Preferred Power and Oil Soill Due to Main Transformer Fault While Shut Down.
y This issue was previously reviewed and documented in Section 08.5 of NRC '
Inspection Report No; 50-293/98-01 with no violations of NRC requirements identified - This LER is closed.
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1 08.2 (Closed) LER 5C-293/98-06-01: Failure to Verifv Comoliance with Technical Specification (TS) Limit for Control Rod Worth.
This LER documented that the control rod withdrawal sequence provided by General '
Electric (GE) did not explicitly verify compliance with the TS limit on control rod worth. Beginning with cycle 7, GE evaluated the dropped rod accident on a generic basis (i.e., based on 280 cal /g) and no longer verified compliance with the plant specific control rod worth limits. Although compliance with the GE rod withdrawal sequence appropriately limits enthalpy and satisfied the purpose of that specification, it did not verify compliance with the maximum control rod worth specified in TS 3.3.B.3. A plant specific analysis was subsequently performed by GE which verified that the TS requirements for rod worth were met for the current operating cycle (cycle 12).
The inspector conduced an on-site review of the LER and verified that the core design procedure was in the process of being revised to include TS 3.3.B.3 to ensure that the cycle 13 core design will meet licensing requirements. The inspector determined that the f ailure to verify the maximum rod worth was a violation of technical specifications. This non-repetitive, licensee. identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-293/98-07-01). This LER is closed.
II. MAINTENANCE M1 Conduct of Maintenance M 1.1 General Maintenance and Surveillance a. Insoection Scooe (61726)
The inspector observed portions of selected surveillance and maintenance activities to verify proper use of approved procedures, conformance to limiting conditions of operation, and correct system restoration following testing. Emphasis was placed on maintenance rule trending and status changes. The following activities were observed:
8 A.15 "HPCI System Integrity Test" 8.5.4.1 "HPCI System Pump and Valve Quarterly Operability" 8.M.2-2.1.10 "4160 Volt Emergency Bus A5 and A6 Loss of Voltage and Degraded Voltege" 3.M.3-7.1 " Replace "B" Motor / Generator (MG) Set Motor Exciter Brushes" 8.9.1 " Emergency Diesel Generator (EDG) Surveillance" MR E9800082 Installation of RHR Loop Decontamination Connections MR 19800473 CRD Pump "B" Casing Gasket Replacement
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b. Observations and Findinas
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!- Mock-up training was conducted at the training facility to f acilitate replacement of the motor exciter brushes on the "B" reactor recirculation system MG set. The l inspector observed that a maintenance supervisor attended the briefing along with l operations personnel. Two electricians replaced the brushes using extra precautions since the MG set was operating. The electricians completed the work well with no problems noted by the inspector. The electricians noted some carbon build-up on l the collector ring and noted the condition in the surveillance procedure. Visual l examination of the brushes by the inspector determined that the degradation was due to normal wear-and-tear. The inspector determined that the use of the training l mock-up and extra electrical work precautions contributed to successful replacement of the electrical brushes.
The inspector observed the maintenance work area drain process for the installation of a flanged connection in the RHR flushing line. The flanged connection will be used as a connection point during the upcoming RHR system piping decontamination. The maintenance area piping was isolated but could not be l drained fully. When the piping was initially breeched, maintenance personnel used tygon tubing to drain water down to the floor in the "B" RHR quadrant room. The inspector observed that the tubing was not well secured in that it was off the floor which caused water to splash in the area and cross a contaminated area boundary.
Licensee maintenance supervisory staff independently observed the condition and corrected the problem and verified that no spread of contamination occurred.
l In preparation to replace the "B" CRD pump casing gasket, the licensee experienced problems with boundary valve seat leakage into the work area. The inspector '
estimated the leakags at approximately one gallon per minute. Maintenance management directed that the work be placed on hold pending repair of the boundary valves. The maintenance staff determined that although pump l disassembly was possible, problems would be experienced during the pump re-
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assembly phase. The inspector determined that the maintenance staff exhibited conservative decision making by placing the work on hold in lieu of attempting to perform the work with boundary valve leakage.
Observed surveillance activities were performed safely and in accordance with approved procedures. Test details were discussed at a pretest briefing with all involved testing participants. During observation of the surveillance activities, proper use of self-checking and procedure adherence was noted. Test personnel l were also noted to demonstrate good radiological protection practices. The surveillance activities were well supported by radiation protection and
- instrumentation and control personnel. The system engineer observed the test and l performed preliminary evaluations of the test results. The inspector verified proper j system restoration and that the test acceptance limits were met. No deficiencies
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6 c. Conclusion
! . Use of mock-up training and extra electrical work precautions contributed to successful replacement of the electrical brushes on the exciter for the "B" MG set motor.
The maintenance staff exhibited conservative decision making by placing the work to replace the "B" CRD pump casing gasket on hold in lieu of attempting to perform the work with boundary valve leakage.
Surveillance testing was done in a well controlled manner consistent with safety requirements. Testing was sufficient in scope to demonstrate that the subject equipment would perform their safety functions. Engineering and radiation protection personnel provided effective testing and radiation protection oversight.
M1.2 Surveillance Freauency a. Insoection Scoce (61726)
The inspector reviewed technical specifications (TS) and a selected sample of infrequently performed TS surveillances in order to verify that the surveillances were completed within the allowable time intervals and properly documented in the master surveillance tracking program (MSTP) database.
b. Observations and Findinos
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The inspector reviewed TS surveillance requirements for the core and containment
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cooling,125VDC, 250VDC, core spray, s'tandby gas treatment, and drywell/ torus l
header systems and verified in the MSTP that their past performance had been accurately tracked and future performances scheduled as required. Additionally, the inspector reviewed copies of surveillances 8.9.8.2 - 125VDC battery acceptance test, 8.9.2.3 - 250VDC battery acceptance test, and 8.7.2.2 - standby gas i treatment inlet heater capability test and verified that the documented performance completion date was accurately reflected in the MSTP.
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The inspector determined that the performance and scheduling of selected infrequently performed TS required surveillances were accurately tracked in the MSTP for the core and containment cooling,125VDC, 250VDC, core spray, l standby gas treatment, and drywell/ torus header systems, i
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M2 Maintenance and Material Condition of Facilities and Equipment M 2.1 Non ASME Code Repair on Salt Service Water (SSW) Pioina a. Inspection Scoce (62707)
The licensee identified leakage coming from SSW system piping spool piece JF29-8-4. The spool piece is located in the SSW system outlet piping from the "A" reactor building closed cooling water system heat exchanger. The leakage consisted of two discrete through-wall pits. The inspector reviewed the engineering evaluation and observed the maintenance repair activity, MR 1981846.
b. Observation and Findinas The inspector verified that the licensee obtained verbcl NRR approval to perform the temporary non-code repair of the SSW spool in accordance with NRC GL 90-05,
" Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1,2, and 3 Piping." The repair consisted of welding a stainless steel metal patch to add strength to the degraded area. A review of the maintenance activity and maintenance records revealed that welder's qualifications were up to date.
Inspection of the repaired area revealed no leakage with the system in service. The inspector noted that the quality assurance review of the work activity revealed that l the wrong filler weld metal was used for the repair. A problem report and non-conformance report (NCR) were issued to document this concern. The disposition of the NCR determined that the metal used was acceptable for the application.
The licensee submitted letter 2.98.109 to the NRC for review and approval of the n'o n-ASME code repair. The 10 CFR 50.59 safety evaluation determined that an unreviewed safety question was not introduced by the temporary pipe repair. Based on operating history, the licensee determined that the cause of the leaks were due to localized delamination of the SSW piping rubber lining. The licensee committed to visually monitor the repaired area once a shift until a permanent ASME code repair is completed, which will be performed during the next scheduled outage exceeding thirty days and no later than startup from the next refueling outage (i.e.
RFO12). In addition, the licensee will perform biweekly monitoring (ultrasonic testing) of the adjacent area and ultrasonic testing of the cover plate material every three months until test results show that the test frequencies can be changed. The licensee inspected nine, vice the required five, similar piping locations in accordance with GL 90-05. These inspections found all locations greater than the manufacturer's tolerances.
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c. Conclusions The inspector determined that the temporary repair of SSW spool r.. ace JF29-8-4 was adequately performed and that the operability determination was appropriate in scope and detail. A thorough review of the work package by quality assurance personnelident!1ied that the incorrect weld metal was used in the repair of the SSW pipa; however, this was determined to be acceptable. The licensee properly utilized i
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[ 8 NRC GL 90-05 to develop and execute the non-ASME repair plan for the degraded SSW spool piece. The inspection of nine similar pipe locations for erosion / corrosion was more rigorous than the GL requirements.
M2.2 (Closed) VIO 98-02-03: Dearaded Eauioment Problem laentification l
The violation related to the NRC identification of several degraded plant equipment conditions which were not known by the plant staff. The licensee determined the reason for the violation was the failure to communicate management's expectations for prompt identification and correction of low threshold material deficiencies. The inspector determined that corrective actions were taken or planned for each individual problem identified in the violation.
Several broader corrective actions were also implemented, including re-emphasis of management expectations and the 12 week rolling maintenance schedule was changed to 13 weeks to dedicate a week for repair of lower level degraded equipment conditions. The inspector determined that the corrective actions taken
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and planned were reasonable and resulted in a large increase in the number of work request tags.
Licensee cross functional teams also performed detailed inspections of designated plant areas to identify and correct adverse equipment conditions. For example, the general material condition of the control rod drive (CRD) quadrant room was significantly improved. Chronic casing leakage was corrected on one CRD pump with repairs planned for the other. The inspector also noted the progress in preparations to conduct on-line chemical decontamination of the residual heat removal (RHR) loops. This was planned to improve the material and radiological conditions in the two RHR quadrant rooms which were presenti designated as contaminated and high radiation areas. The chemical decontamination is planned in the fall 1998. Lastly, the inspector noted the installation of permanent lighting in the torus room. This violation is closed.
Ill. ENGINEERING E2 Engineering Support of Facilities and Equipment E2.1 (Closed) LERs 50-293/98-08 and 98-16: (Closed) Item Ill.D.3.4 of NUREG 0737 (Ocen) VIO 50-293/98-02: Control Room Hiah Efficiencv Air Filtration System
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a. inspection Scope l
l The inspector performed a walkdown of the control room high efficiency air filtration system (CRHEAFS) and examined system operating, surveillance and alarm response procedures. Design calculations for the CRHEAFS were also reviewed along with the system description contained in the plant Final Safety Analysis Report (FSAR).
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l 9 The CRHEAFS consisted of two redundant subsystems; a non-safety related portion used during routine plant operations, and a safety related portion that would be manually activated during a Loss of Coolant Accident (LOCA). The non-safety related subsystem was designed to maintain the temperature of the control room environment within desirable values, by passing a mixture of fresh and recirculated -
air through a series of heating and cooling coils. When activated during a LOCA, the safety-related subsystem was designed to prevent the infiltration of radionuclides by l pressurizing the control room envelope above atmospheric pressure by supplying l outside air through charcoal filters. Each subsystem had redundant components, j
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i.e., there were two safety related charcoal filters and fans along with two non-safety related heating / cooling coils and supporting fan units. Both systems share some components; for example, both subsystems used the same air intake and exhaust plenums.
b. Observations and Findinas Desian inadeauacies l In April, the NRC identified that some portions of the safety related, seismic category 1 control room ductwork subsystem were suspended from the non-safety related, seismic category 2 ductwork by metal straps. The licensee documented this condition in Problem fleport 98-9202. Initially the licensee determined that this configuration was acceptable. A follow up detailed review completed on May 8,
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1998, determined the arrangement was a condition that was outside of the plant l
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design basis and additional modifications may be necessary to achieve full seismic qualification. This conclusion was reported to the NRC operations center as required by 10 CFR 50.72(b)(1)(ii)(B). The corrective action as outlined in Licensee Event Report (LER) 50-29'3/98-08, dated May 27,1998, included a commitment to perform additional analysis and modifications as necessary to assure full seismic qualification.
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The inspector conducted an on-site review of LER 98-08 and determined it contained an accurate description of the issue and corrective actions were adequate. Therefore, this LER was closed.
10 CFR 50, Appendix B, Criterion lil, " Design Control," requires that " measures shall be established to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into o specifications, drawings, and procedures." The failure to have the as-built configuration of the CRHEAFS system in accordance with the design basis indicated inadequate measures were established to translate the design basis of the CRHEAFS into procedures, instructions and drawings. This is the first of 3 examples of a violation (VIO 50-293/98-07-02)of 10 CFR 50, Appendix B, Criterion 111, " Design Control."
Subsequent licensee reviews of the accident mode testing of CRHEAFS identified 4 additional seismic deficiencies and a single failure vulnerability. Specifically, the l licensee noted if damper AO-N 1 and its supporting ductwork in the non-safety
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related subsystem failed to open during a LOCA or seismic event, outside air could enter the control room environment without going first through the charcoal filters located in the safety related subsystem of CRHEAFS. . As a result, the radiation dose operators would receive during a LOCA may increase, or the control room environment may become uninhabitable following a seismic event.
When this deficiency was detected, the licensee declared both trains of the CRHEAFS inoperable and entered the 36-hour shutdown action statement contained in Technical Specification (T.S.) 3.7 " Control Room Air Filtration System."
Corrective action as outlined in LER 50-293/98-16 consisted of installing seismic restraints on the non-safety related portion of the seismic ductwork and gagging closed the damper. Once the work was completed, the TS action statement was exited.
The inspector conducted an in-office review of LER 98-16 and determined it contained an accurate description of the issue and corrective actions were adequate. Therefore, this LER was closed.
Chapter 10.17 of the Pilgrim Final Safety Analysis Report (FSAR) states that the CRHEAFS was designed with sufficient redundancy so that no single active failure could prevent the system from achieving its safety objective. However, the as-built system configuration was not in accordance with this design requirement because i one active single failure, the failure of damper AO-N-1 to close, could render the CRHEAFS inoperable during a LOCA. This represents the second example of a violation (VIO 50-293/98-08-02)of 10 CFR 50, Appendix B, Criterion 111, " Design Control."
Analysis Weaknesses In a letter dated February 11,1981, the licensee provided the NRC part of its response to item lli.D.3.4, Control Room Habitability of NUREG 0737, " Clarification of TMI Action items," including the results of an analysis, which indicated the
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CRHEAFS met the exposure limits contained in GDC 19. The analysis results were reviewed and accepted by the NRC in a letter dated June 24,1982.
The inspector reviewed portions of the February 11,1981, submittal and determined some assumptions in the analysis did not have an adequate basis and others may no longer be valid. Specifically, the analysis did not appear to account for the time delay that would occur between receipt of a high radiation signal in the control room ventilation intake and the manual operator action to shut down the
normal ventilation. The time delay could be considerable, because stopping the l non-safety related subsystem requires manipulation of switches outside of the l control room envelope at elevation 51 of the turbine building.
l The isolation delay was further affected by the alarm response procedure for a high radiation signal in the control room ventilation intake, which instructed operators to first confirm the signal was valid before isolating the intake. This delay could be considerable because signal confirmation involved having a health physics
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I technician confirm the signal was valid by taking radiation surveys with a meter at
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- Because of these analysis and procedure deficiencies, during a LOCA operators
could be exposed to radiation levels that were greater than contained in the plant
- design trasis. The failure to transfer the control room habitability analysis into the
CRHEAFS operational procedures reoresented the third example of a violation (VIO j. 50-293/98-08-02)of 10 CFR 50, Appendix B, Criterion ill, " Design Control."
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At the close of the inspection report period, the licensee corrected the alarm a response procedure to specify that operators initiate CRHEAFS immediately when ,
- high radiation was detected in the control room intake, in addition, an engineering - 1 evaluation was performed that concluded that the dose to control room personnel l would remain below the GDC 19 limits assuming a time delay of 30 minutes post i
!- receipt of the radiation alarm for CRHEAFS initiation. During a September 11, j
1998, telephone call, licensee engineering management stated that to identify and j correct other possible system design errors, analysis weaknesses, and procedure ;
- inconsistencies, a design basis information review effort of the CRHEAFS would be i j performed.
Based on the review of the engineering evaluations, corrective actions taken, and a
walkdown of the CRHEAFS, the CRHEAFS remains operable and capable of l performing its design function. Therefore, item Ill.D.3.4 of NUREG 0737 is !
j considered closed.
l c. Conclusions i The NRC and the licensee's engineering staff identified deficiencies in the design of the control room high efficiency air filtration system, i.e., seismic qualification of duct supports, single failure design analyses, and inconsistencies between analysis assumptions and operational procedures. The deficiencies have been reported in two LERs, including acceptable short term corrective actions. The failure to accurately transfer the system design into instructions, procedures or drawings was a violation of 10 CFR 50 Appendix B Criterion til " Design Control."
E2.2 (Closed) VIO 98-02-07:Inadeauate Desian Control l The violation related to self disclosing events in late 1997 when the "A" SSW pump motor shaft failed twice. The cause was human error when engineers performing a substitution equivalency evaluation did not include the actual configuration of the pump pedestal which had been previously modified. As a result, motor misalignment was created that resulted in shaft failure. The maintenance staff returned the pump pedestal to its original configuration and replaced the shaft. No further alignment problems were experienced. Engineering management discussed this event with all engineering personnel to review the lessons learned. This violation is closed.
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l E8 Miscellaneous Engineering issues (92903)
E8.1 (Closed) IFl 97-02-03: Jet Pumo Nonconformina Conditions l l
! During inspections of the reactor core internals during the last refueling outage (i.e.,
l RFO11), the licensee identified that several jet pump swing gate latch pins were not I l fully engaged and some restrainer bracket set screws were not in contact with the i
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jet pump mixer. The licensee hired a consultant who completed a detailed structural analysis of the jet pumps. This evaluation concluded that the structural integrity I was assured during design basis conditions. An NRR safety evaluation is attached to this report as Attachment 1. As stated in the NRR evaluation, the licensee plans l
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to conduct additional jet pump inspections during RF012 to confirm that the l aforementioned conditions have not changed. The inspector confirmed through I discussions with licensee engineering managers that they are still planning to l conduct this inspection. The inspector determined that the licensee met the intent of NRC Generic Letter 91-18 during resolution of this issue. This item is closed.
IV. PLANT SUPPORT P3 EP Procedures and Documentation
- A number of specific revisions to EP procedures, listed in Attachment 2 to this inspection report, were examined during an in-office review. Based on the licensee's determination that the changes do not decrease the overall effectiveness
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of the emergency plan and after limited review of the changes, we determined that no NRC approval is required. implementation of these changes will be subject to
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inspection in the future to confirm that the changes have not decreased the overall l effectiveness of your emergency plan.
l S1 Conduct of Security and Safeguards Activities !
l S 1,1 Control of Soecial Nuclear Material a. Inspection Scope (71710)
The inspector reviewed the circumstances surrounding the July 29,1998, report of missing special nuclear material (SNM). Four intermediate range monitors (IRMs)
were not able to be located during a semi-annual audit of SNM.
b. Observations and Findinas The IRMs had been removed from the reactor core and according to the material
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, transfer forms were placed in the spent fuel pool (SFP) in a container in the November-December 1997. Each IRM contains approximately 0.008 grams of total uranium, which has approximately 0.05 microcurie of radioactivity.
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The licensee executed a search plan of the spent fuel pool and placed a hold on the removal of the radioactive waste storage facility. The licensee searched the immediate area of the SFP and moved several fuel bundles in an attempt to locate the container. The search resulted in the discovery of two of the four IRM detectors and several pieces of LRM cabling. A search of the radioactive waste storage facility resulted in the location of the remaining two IRMs.
During discussions with the reactor engineering supervisor he stated that the apparent cause for the misplacement of the SNM was personnel error and insufficient care in the movement / storage of material. During a forced shutdown in December 1997, a number of IRM detectors were replaced. When transferring the IRMs to the SFP, technicians noticed that several pieces of IRM cable were mixed in with the detectors. This material was removed and the detectors were placed in the SFP and the cable pieces were transferred to the radwaste storage facility. An error occurred at this poirit and 2 of tha 4 IRM detectors were sent to the radwaste storage facility. The engineer in charge of the transport of the SNM did not verify removal of the IRMs from the bucket to the SFP. The engineer was under the assumption (based on the pre-job brief) that the transfer bucket was to contain only four IRMs. The engineer verified, from the SFP bridge, that four items had been transferred from the transfer bucket to the SFP storage bucket. The instrumentation and control technician (l&C) performing the transfer failed to properly verify the transfer of the IRMs. The inspector determined that the failure of the licensee to properly control the IRM material transfer is a violation of 10 CFR 70.51(b)(1).
Corrective action planned for this issue include: changing the rr.sthod of removal of IRMs from the core to minimize the need to identify detectors from cabling, train l&C personnel on the event, and change the method of storage of IRMs to prevent loss of material should the storage container be mishandled. The inspector reviewed root cause evaluation and corrective actions and determined that they were acceptable. This non-repetitive licensee identified and corrected violation is being treated as a Non-Cited Violation (NCV 50-293/98-07-03), consistent with Section Vll.B.1 of the NRC Enforcement Policy. The inspector verified that the licensee had considered the reportability criteria delineated in 10 CFR 70.52 and 10 CFR 20.2201. No NRC notifications were required.
c. Conclusions The circumstances leading to the misplacement of the intermediate range detectors represented a weakness in the accountability and control of special nuclear material.
The licensee was successful in locating the missing intermediate range monitors.
The corrective actior s to resolve the issue were determined to be appropriate.
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i V. MANAGEMENT MEETINGS
. X1 Exit Meeting Summary l The inspector presented the inspection findings to members of the licensee management after conclusion of the inspection on September 29,1998. The l licensee acknowledged the findings presented.
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INSPECTION PROCEDURES USED l
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! IP 37551: Onsite Engineering i. IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observation IP 62707: Maintenance Observation IP 71707: Plant Operations
- IP 71750
- Plant Support Activities l lP 82301: Evaluation of Exercises for Power Reactors l IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor
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Facilities IP 92901: Followup - Operations ,
IP 92902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: Followup - Plant Support i
' IP 93702: Prompt Onsite Response to Events at Operating Power Reactors
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!. 16 l l ITEMS OPENED, CLOSED, AND UPDATED
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i VIO 50 293/98-07-02 Control Room High Efficiency Air Filtration System !
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I Closed l LER 50 293/97-04-01 Loss of Preferred Power and Oil Spill Due to Main Transformer Fault While Shut Down.
!~ LER 50-293/98-06-01 . Failure to Verify compliance with Technical Specification (TS) l Limit ;
- LER 50-293/98-08 A Section of CRHEAFS Seismic Class 1 Ductwork found with !
( Class ll '
LER 50-293/98-16 Control room Ventilation Exhaust Damper and Ductwork not I Seismically Qualified j NCV 50-293/98-07-01 - Failure to Verify compliance with Technical Specification (TS) l l Limit i NCV 50-293/98-07-03 Loss of Special Nuclear Material i l IFl 50-293/97-02-03 Jet Pump Nonconforming Conditions i VIO 50-293/98-02-03 - Degraded Equipment Problem Identification !
VIO 50-293/98-02-07 Inadequate Design Control l
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, LIST OF ACRONYMS USED t
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AOT Allowed Outage Time ASME- . American Society of Mechanical Engineers
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BECo Boston Edison Company CFR Code of Federal Regulations CRD Control Rod Drive CRHEAFS Control Room High Efficiency Air Filtration System L
EDG' ' Emergency Diesel Generator FSAR Final Safety Analysis Report '
L 'GDC. General Design Criteria l GE General Electric L
'HPCI High Pressure Coolant injection IFl. -Inspection Follow-Up Item i
. lR ' in.spection Report
- IRM - Intermediate Range Meters
- LER. Licensee Event Report LOCA Loss of Coolant Accident
- MG- Motor Generator MR.; Maintenance Request = !
' MSTP Master Surveillance Tracking Program ,
l- NCR Non-conformance Report
~NCV ..Non-Cited Violation NOV Notice of Violation NRC- Nuclear Regulatory Commission NRRL Office of Nuclear Reactor Regulation "
! PNPS ~ Pilgrim ' Nuclear Power Station PR Problem Report RHR: Residual Heat Removal *
p SBGT- Standby Gas Treatment
- SFP_ . Spent Fuel Pool-
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i_ - SSW.. Salt Service Water L :TS Technical Specification UFSAR- Updated Final Safety Analysis Report. t
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VIO Violation p JWWM Work Week Manager i
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ATTACHMENT 1 TIA RESPONSE REGARDING DIFFEacijCES IN THE "AS DESIGNED" AND "i,S FOUND"
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CONDITIONS AT THE PILGRIM NUCLEAR POWER STATION
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BACKGROUND !
l During the April 1995 Refueling Outage (RFO) 10, the jet pumps at the Pilgrim Nuclear Power Station (PNPS) were inspected for evidence of tack weld cracks as recommended j by General Electric (GE) Services Inforrnation Letter (SIL) No. 574. While no weld cracks ,
were identified, gaps were observed between the set screws and inlet-mixers on several l pumps. The original jet pump design required that the set screws be in contact with the inlet mixers. An operability evaluation performed at the time of the inspection concluded '
that no corrective action was required prior to startup.
In June 1996, GE apprised the owners of BWRs that jet pump restrainer bracket set screw l gaps had been observed in a number of GE BWRs, and could be present in any BWR with l jet pumps. The GE Rapid information Communication Service Information Letter (RICSIL)
l No. 078 advised that, "These gaps reduce the pump lateral support near the bottom of the '
jet pump inlet-mixer and increase the potential for flow induced vibration of the inlet-
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mixer." The GE communication noted that, "If significant set screw gaps are present, a plant specific analysis may be required, and has been performed for some pumps, to determine the acceptability of continued operation." The RICSIL turther advised that, at !
one BWR, gaps had been observed to form on one pump, increase in size on another J
pump, and decrease in size on another pump over one operating cycle; and that the root cause of how the gaps are formed. is not known. .
While preobring for a vessel inservice inspection during the February 1997 RFO-11, a i review of RFO-10 video tapes revealed incomplete engagement of the swing gates on jet !
pumps numbers 5 and 11. This discovery resulted in the inclusion of visual inspections of all jet pump swing gates and set screw gaps in the RFO-11 inspection scope.
The ingactions conducted during RFO-11 revealed that some of the swing gate to I rastrainer brr cket latching pins were not fully engaged in the latch pm coles. Only two new swing gates, for pumps 5 and 11, had been procured for the outage, and the discovery of the remaining incomplete latched swing gates left no time to obtain additional j replacement swing gates. However, the gates were secured to the restrainer bracket by
, two captured and torqued gate keeper bolts. Also, various jet pumps did not have their l restrainer bracket set screws in full contact with the inlet-mixer. Boston Edison Company l (BECo) completed a Safety Evaluation (SE) that concluded that it was acceptable to leave the conditions "as-found." BECo's SE included qualification analyses, performed by Structural Integrity Associates, Inc. (SI), as recommended in GE's RICSIL No. 078.
REVIEW REQUEST
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By letter dated April 22,1997, Region I, Division of Reactor Projec s, forwarded the subject Task Interface Agreement (TIA) and requested NRR's assis;tance in determining th acceptability of BECo's conclusion that no unreviewed safety question (USQ) existed as a result of differences in the "as designed" versus "as founo" conditions in the jet pumps at PNPS. BECo piepared SE No. 3084 dated April 2,1997, regarding the differences between the "as designed" and "as found" features of the jet pumps. The Region requested that NRR's review of the SE include an assessment of the schedule and extent of appropriate future inspections of the jet pump affected areas.
REVIEW
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i Jet pumps are part of the reactor coolant recirculation system and are located in the annulus region between tha core shroud and reactor vessel wall. PNPS has 20 jet pump assemblies arranged in pairs. Each jet pump assembly pair consists of a common riser pipe, a riser brace, a transition piece, two inlet mixer assemblies, two inlet nozzles, and two tailpipes connected by an integral restrainer bracket. The inlet mixers exit into separate, non-integral diffuser assemblies. The riser pipe is connected to the recirculation inlet nozzle thermal sleeve by way of a 90 degree short radius elbow. The riser pipe is supported near the inlet mixer by the riser brace which is welded to the riser pipe and to pads attached to the reactor vessel wall. The entrance end of each inlet mixer assembly is clamped mechanically to the riser transition piece by a bolted beam assembly. The restrainer brackac assembly provides lateral stiffness to a pair of jet pump inlet mixer assemblies. The restrainer bracket assembly provides a nonintegral three point connection for each pair of inlet mixer tailpipes. Each restrainer assembly consists of a stationary bracket welded to the riser pipe and a clamp arrangement made up of a swing gate and
latch mechanism which encloses an individual intet mixer tailpiece. Two set screws are *
attached to a bracket and, through :ontact vJith the inlet mixer, adjust the position of the inlet mixer tailpiece with respect to the diffuser. The swing gate is latched to the restrainer bracket by a pin and is secured to the bracket by two gate keeper bolts pretensioned to 40-45-foot pounds. The swing gate is provided with an adjustable wedge which assures the three point contact between the inlet mixer tailpiece and restrainer. Instrument lines are attached to the jet pumps to sense jet pump flow and provide indication of recirculation flow mismatch and water level monitoring. The jet pump assembly, although a non-ASME Code component, is safety-related because the structural ir,tegrity of the assembly assists in maintaining core refloodability following a postulated Loss of Coolant Accident (LOCA)
and thus, is essential for safe shutdm m.
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BECo contracted Si to perform analyses to assess the jet pump frequency response and structural impact changes resulting from differences between the "as designed" and "as found" conditions. The purpose of the analyses was % determine whether the set screw to inlet mixer gaps and inrrmplete engagement of the restrainer bracket swir.g gate latches would have a detrimental effect on the operation of the jet pump w;th regard to increased j vibration or structural degradation.
i The original design and analyses of the PNPS jet pump assembly could not be locateriin ,
.BECc's er GE's archives. GE provided design basis loading conditions and a new set of !
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seismic response spectra based on the recent re-analysis of the reactor vessel shroud. The loading conditions were reconstructed and updated to reflect current criteria based on a review of the FSAR, drawings, and the reactor vessel analysis of record.
The Si analyses were based on a two-dimensional, jet pump finite element model representing the piping, elbows, inlet mixers and diffusers. The model was fixed at the recirculation inlet nozzle end laterally and vertically restrained at the riser brace by a flexible member attached to the pressure vessel wall. The restrainer bracket assembly was modeled as a series of massless beams that linked the riser pipe to the inlet mixers to restrict horizontal motion but allow rotation and vertical movement. The model represented the "as designed" condition. The "as found" condition model removed the beams that linked the riser pipe to the inlet mixer. Analyses were performed using the above described model to determine the effects on jet pump operation under "as found" conditions as compared to "as designed" conditions.
The results of the frequency analysis show that the "as found" condition with set screw gaps seduces the fundamental frequency of the pump assembly to 26.4 Hz as compared to the original design fundamental frequency of 34.3 Hz. Both frequencies are not within the region of resonance with the pump blade passing frequency of 139 Hz or a primary harmonic or sub-harmonic of that frequency.
The structural analyses performed by Si compared the "as designed" and "as found" conditions with regard to differences between load distribution, stress, and deformation.
The structural evaluation determined that the structural integrity of the jet pumps with set screw gaps and a dislodged restrainer bracket swing gate remains within acceptable l loading and the specified stress limits. All reported stress limits are within the ASME Code allowable limits. The analyses show that the stress levels for the "as designed" and "as foun'd" conditions are very similar. The stress levels and' fatigue usage factors for the jet pumps remain effectively unchanged. Si performed analyses on the capability of the two gate keeper bolts, torqued to 40-45 foot pounds, to hold the swing gates in place. Sl concluded that an applied torque of 40 foot pounds is adequate for design conditions.
BECo's SE concluded that, as a design consideration, the gate bolt torque will provide adequate restraint for normal and transient operating conditions.
BECo's SE summary concluded that, based on the Si analyses arid the RFO-11 inspections, I the structural integrity is assured under design basis conditions, and the impact of I vibratory loads are minimal as confirmed by 10 years of inservice inspections. No )
indications of adverse jet pump motions were identified during the RFO-11 inspections, l which is consistent with the analysis demonstration of minimalinlet mixer deflections.
Therefore, inservice loadings have been ruled out, by BECo, as the cause of the set screw gaps.
EVALUATION
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l The staff reviewed the information provided in BECo's SE, including the attached Si analyses and report of the RFO-11 inspection results. Based on the information provided, j and in consideration of BECo's report that periodic in-vessels inspections over the past five fuel cycles have not detected any fatigue cracking, excessive wear indications, or evidence of damage to the jet pumn components, the staff believes that BECo's conclusion that no USOs exist is acceptable.
With regard to future inspections, we recommend that the following considerations be addressed by the licensee within the scope of inspection of jet pump's components affected by the as-found conditions:
1) Based on findings from the RFO-11 inspection of the jet pumps, BECo developed an explanation for the cause of the incomplete swing gate latching that apparently resulted from an incorrect installation performed on the gate replacement for jet pumps 5 and 11. BECo concluded that the root cause of the incomplete latching was the result of either improper installation, or improper fabrication, and that the condition had existed for 10 years. It was further postulated that this provided a
, reasonable explanation for the lack of metal-to-metal contact between the restrainer
bracket set screws and inlet mixers. Although the explanation appears reasonable, except for the experience gained during RFO-11, no evidence from any previous inspections has been provided to verify that the proposed root cause has existed for over 10 years. Further, RICSIL No. 078 reports that at one BWR, gaps have been observed to form on one pump, increase in size on another pump, and decrease in size on yet a third pump over one operating cycle.
2) The lateral restraining force which prevents unrestrained inlet mixer motion is attributed to the swing gate torqued keeper bolts. The torqued keeper bolts provide
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a compressive clamping force that resists motion. This clamping force exerts a lateral friction force proportional to the compressive force and the selected coefficient of friction. The friction force is calculated under static load conditions.
In a vibratory environment the contact area of the compressive force could be subject to wear, which over time would reduce the compressive force and change the coefficient of friction.
The licensee's SE indicated that, over the past 10 years, the data obtained during the RFO-11 inspection represented the only reliable and accurate information on the set screw gap measurements, assessment of the swing gate latching conditions, and the torqued keeper bolts integrity. On this basis, we recommend that the licensee reinspect a sufficient portion of these areas and compare the results with the RFO-11 data as a baseline to verify and support its contention that the conditions have not changed and have existed for over 10 years.
@NCLUSIONS
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We conclude that the licensee's determination that no USO exists is acceptable. Further, !
our view with regard to the scope and interval of appropriate future inspections, is that the s
.li censee should reinspect sufficient portions of the "as found". conditions of the jet pumps at an interval to be developed and justified, to verify and support their contention that ,
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conditions have not changed and have existed over the past 10 years. Results of future ,
inspections should be compared to data from the RFO-11 as a baseline. i t
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23 ATTACHMENT 2 >
Emergency Plan implementing Procedures Reviewed Procedure l Number Title Revision No.
- -210 ' Control Room Augmentation SA
- -229 TSC/OSC Equipment Operation 4
- -230 OSC Activation and Response 1A
- -231 Onsite Radiation Protection 4 l
- -240 - Emergency Security Organization
- Activation and Response 7
- -250 EOF Activation and Responsa 6A
- -251 Offsite Radiation Protection 3 '
- -252 Facilities Support - 5
- 254 Communications Support 1 .
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! #-310 Radiation Monitoring Team Activation and Response 2
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- -315 Offsite Personnel Monitoring Team Activation and Response 3
- #-330 Core Damage 2B
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- -410 ~ Evacuation / Assembly 3
- -420 Search and Rescue 1A
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- -440 Emergency Exposure Cor.trols 5A l
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