IR 05000336/1990006
| ML20043G329 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/11/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20043G327 | List: |
| References | |
| 50-336-90-06, 50-336-90-6, IEB-88-002, IEB-88-2, NUDOCS 9006200159 | |
| Download: ML20043G329 (28) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.:
50-336/90-06 Docket No.:
50-336 License No.
DPR-65 Licensee:
Northeast Nuclear Energy Company h.0. Box 270 Hartford, CT 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2
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Inspection at: Waterford, Connecticut Dates:
March 6 - April 16, 1990 Reporting Inspector:
Peter J. Habighorst, Resident Inspector Inspectors:
Wi yam J. Raymond, Senior Resident Inspector P e ighorst, Resident Inspector
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[7/// 98 Approved by:
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TTonald R. F(ver a
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/ pte eactor Prc1 s ection 4A Division of Reac or Projects
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InspecthonSummary: Inspection on March 6, 1990 - April 16, 1990, Inspection heport No. 50-336/90-06 Areas Inspected:
Routine NRC resident inspection of plant operations, surveillance, maintenance, previously identified items, engineering / technical support, committee activities, and security events.
Results:
1.
General Conclusions on Safety perspective, Strengths or Weaknesses in Licensee Programs At the end of this inspection period, the licensee confirmed a primary-to-secondary leak existed in one steam generator (SG), with leakage values that were a small fraction of the technical specification (TS) limit. While some uncertainty existed regarding the exact leakage mechanism, the major indications were that an indirect leak path existed rather than a more significant degradation in an inservice tube, hhk D
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Significant licensee management and control room operator awareness of leakage values and trends were noted.
Licensee actions to monitor the leakage is considered good based on initiatives to quantify low levels of h
leakage, implement additional monitoring capabilities not required by TS, use engineering support to assess the steam generator status, and the use of additional chemistry evaluations (i.e., oredictive tritium analysis).
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Violations
Within the scope of this inspection, no violations were observed.
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Unresolved Items Two previously unresolved items were closed (sections 3.6 and 5.3.1).
Two items were opened during the report period, regarding; (1) the need to update FSAR cable separation information to agree with 10 CFR 50 Appendix R Section III.G. exemptions (section 3.5) and; (2) a non-conservative technical specification for pressurizer cooldown rates (section6.1).
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Executive Summary Plant Operations (Module 71707/93702/64704) The licensee has implemented a good monitoring program to trend a primary-to-secondary leak in one steam gen-erator. Licensee initiatives to monitor and assess the leak rate were good.
Additional licensee actions to further characterize the leakage due to un-t..
certainties in the leak mechanism were good and are prudent to assure continued safe operation of the facility.
In the Fire Protection area, NRC inspection identified the need to update the FSAR to make it refleet Appendix R exemptions for cable separation requirements.
Radiological protection (Module 71707/92701)
Facility radiation exposure was very low for the nionth of March (less than one man-rem).
Unit management used daily planning meetings effectively to heighten awareness of ALARA controls in daily work activities.
Surveillance and Maintenance (Modules 61726/62703/92702) Review of routine maintenance identified o noteworthy findings.
Surveillance activities were performed in accord **-
with established test controls.
NRC review of a procedure to cai R
.he reactor building closed cooling water system radiation monitor
,a the test method acceptable.
Security (Module 71707) Routine review in this area identified no noteworthy findings.
Engineering and Technical support (Modules 92702) The licensee identified I
that the technical specification for the pressurizer cooldown rete was in error.
Licensee evaluation of historical data for 5 instances in which the cooldown rate was in excess of the analyzed value identified no unsafe condi-tions, in that a bounding calculation for a similar pressurizer shows an acceptable fatigue analysis result.
The communications between the site and engineering on the issue could have been more timely.
Safet Assessment / Quality Verification (Modules 30703/40500/90712/92702/
927 Review of routine reports identified no noteworthy findings. NRC inspection found I&C supervisor actions responsive to a concern raised by an employee.
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TABLE OF CONTENTS Page 1.0 ' Persons Contacted...............................................
2. 0. Summa ry o f _ Fa c i l i ty Ac ti v i t i e s..................................
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3.0 Plant Operations (IP 71707//93702/64704)........................
3.1 Control Room Observations..................................
3.2 Plant Tours................................................
3.3 Review of Plant Incident Reports..........................
2-3.4 Prima ry to Secondary Leakage Indications..................
3.5 Cable Raceway Separation...................................
3.6 Previously Identified Items................................
3.6.1 (Closed) Unresolved Item 89-22-02: Evaluation of Fire Consequence With an Open Fire Barrier..........
4.0 Radiological Controls (IP 71707/92701)..........................
9-4.1 Posting and Controls of Radiological Areas.................
5.0 Maintenance / Surveillance (IP 62703/61726/92702).................
5.1 Observation of Maintenance Activities......................
5.2 Observation of Surveillance Activities.....................
5.2.1 Review of SP 2404AW for Calibration of Radiation Monitor RM 6038.....................................
5.3 Previously Identified Items................................
5.3.1 (Closed) Unresolved Item 88-24-01: Incorporation
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of Material Equipment Parts List Quality Assurance
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Indicators into the Production Management Maintenance System..................................
6.0 Engineering / Technical Support (IP 92702)........................
6.1 ' Pressurizer Cooldown Rate Di screpancy......................
6.2 Reactor Fuel Performance...................................
7.0 -Security (IP 717107)............................................
8.0 Safety Assessment / Quality Verification (IP 30703/40500/90712/92702/92700)..............................
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8.1 Committee Activities.......................................
8.2 Nuclear Review Board.......................................
8.3 Periodic Reports...........................................
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TABLE OF. CONTENTS l
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Page 9.0 ManagementMeetings(IP 30703).................................
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9.1 Corrections from Previous Inspection Reports.............. 17-
9.2 Concerns Referred to -the Licensee for Resolution..........18'
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9.3 Concerns Reviewed by the Inspector -( A.45.01).............. 18 I
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Attachment A: Allegations Presented to Licensee Management for Disposition l
Attachment B: Review of Technical Concerns on Calibration of Monitor RM 6038 i
The NRC inspection manual. inspection procedure (IP) or_ temporary instruction
(TI) that was used as inspection guidance is listed for each applicable report
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section.
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DETAILS
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L 1.0 persons Contacted
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Interviews and discussions were conducted with licensee staff and
management during the report period to obtain information pertinent to i
t the areas inspected..
L Inspection findings were discussed periodically with the supervisory and l
management personnel identified below.
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S. Scace, Nuclear Station Director, Millstone Station
- J. Keenan, Nuclear Unit Director, Millstone Unit 2
J. Riley, Maintenance Manager, Millstone Unit 2
'J. Becker, Instrument and Controls Manager, Millstone Unit 2
J. Smith, Operations Manager, Millstone Unit 2
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- Attendee at post-inspection exit meeting on April 27, 1990.
2.0 ' Summary of Facility Activities Millstone Unit 2 operated at rated thermal power throughout the inspection period.
On March 29,.the licensee identified a seawater leak in the
'B' main
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condenser.
The following day, the main condenser water box was taken out of service to nondestructively examine the condenser tubes.
The examina-
.tions revealed three tubes with thru-wall defects. The licensee's pre-liminary review identified the potential root cause of tube leakage as i
debris from an extraction steam boot for the '5A' low pressure heater.
Licensee monitoring of heater performance was still in progress at the end -
of the insoection period.
Licensee actions will be followed in routine
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inspections.
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NRC Activities A region I mid-cycle systematic assessment of licensee performance (SAlp)
inspection was conducted between April 3-4, 1990.
Results will-be pro-vided in inspection report 50-336/90-80.
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The inspection activities during this report period included 154 hours0.00178 days <br />0.0428 hours <br />2.546296e-4 weeks <br />5.8597e-5 months <br /> of inspection during normal working hours.
In addition, the review of plant
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operations was routinely conducted during periods of backshifts (evening shif ts) and deep backshif ts (weekend and midnight shifts).
Inspections were performed during 13 backshift hours and during 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of deep
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backshift.
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3.0 Plant Operations (IP 71707//93702/64704)
3.1 Control Room Observations
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Control room instruments were observed for correlation between channels, proper functioning, and conformance with Technical Specifications. Alarm conditions in effect and alarms received in
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the control room were discussed with operators, The inspector-periodically reviewed the night order log, tagout log, Plant
Incident Report (PIR) log, key log, and bypass jumper log.
Each of the respective logs was discussed with operation department staff.
No inadequacies were noted.
3.2 Plant Tours t
The inspector observed plant operations during regular and backshift l
tours of the following areas:
Control Room ESF Cubicles Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure Enclosure Building During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication and equipment status. No inadequacies were noted.
3.3. Review of Plant Incident Reports (PIRs)
The plant incident reports (PIRs) listed below were reviewed during the inspection period to (1) determine the significance of the
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events; (ii) review the licensee's evaluation of the events; (iii)
verify that the licensee's response and corrective actions were
proper; and,-(iv) verify that the licensee reported the events in
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accordance with applicable requirements, if required.
The PIRs reviewed were:
90-13, 90-14, 90-17, 90-20, 90-23, 90-27, and 90-28.
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No inadequacies were noted.
3.4 Primary to Secondary Leakage Indications This part of the inspection focused on primary to secondary leakage history, the condition of the steam generators, off-site dose assessments, procedural requirements for leakage, and overall assessment of licensee actions concerning steam generator integrity.
Technical Specification 3.4.6 provides the limits for operation with steam generator leakage.
The licensing basis for limiting leakage from a single steam generator to 0.10 gallons per minute (gpm), or i
144 gallons per day (gpd), provides assurance of steam generator
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tube integrity and retention of structural margins recommended in NRC Regulatory Guide 1.121 " Bases for Plugging Degraded PWR Steam-Generator Tubes." The total steam generator tube leakage limit of 1.0 gpm ensures dosage contribution from tube leakage will be
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limited to a small fraction of the 10 CFR 100 offsite limits.
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History of primary to Secondary Leakage The licensee's calculation of primary to secondary leak rate is
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primarily prescribed and implemented in procedure CP-2806Y
" Calculating Primary to Secondary Leak Rate." The calculation
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employs two methods:
(1) comparison of isotopic activities of the
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steam generator condensed liquid to reactor coolant; and, (2)
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comparison of air ejector activity to reactor coolant activity. The isotopic ratios are multiplied by the known steam generator (SG)
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blowdown rate or the steam jet off gas rate, and corrected for decay times of specific analyzed isotopes. An additional method for leakage measurement-is comparison of primary and secondary tritium activities as developed in procedure SP-2842 " Unit 2 Condenser Air Ejection Sump (Analysis) for Principle Gamma Emitters and Tritium."
During the time interval from December 1, 1989 to April 7, 1990, steam jet air ejector (SJAE) total activity increased by a factor of approximately one hundred. The associated reactor coolant flash gas
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activity increased over the same period of time by a factor of two.
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However, at the end of the inspection period, the total SJAE activity was declining to a level ten times the value in December,
1989.
The facility experienced an abrupt increase in measured leakage between April 7 - April 9.
Previous to this time calculated leakage rates averaged 5-10 gpd. The leakage between April 7-9 peaked at approximately 35 gpd and decreased to 17 gpd on April 9.
Licensee actions during this time included: (1) increased sampling frequency of the SG blowdown and SJAE from twice per day to twelve times per day, and additional characterization of the isotopic half-lives to quantify time delays of the transport mechanism.
On April 11, the NRC held a conference call with the licensee to discuss the following topics:
status and trending of current leak rates in the steam generators; methods of determining leak rates, and the licensee's response to NRC Bulletin 88-02 " Rapidly Propagating Fatigue Cracks in Steam Generator Tubes " The licensee presented their results for: the SJAE gaseous radioactivity trends; leak rate calculations; postulated primary to secondary leakage paths based on
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measured time delays in the transport mechanism; responsiveness of radiation monitors used to monitor steam generator leakage, including the Nitrogen-16, main steam line and the blowdown monitor; i
steam generator liquid samples during the air ejector activity build-up on April 7; and, use of tritium analysis to confirm the magnitude of the leakage rat,
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As of April 18, the licensee was still evaluating their actions on NRC Bulletin 88-02. These actions will be followed up in future
routine resident inspection.
Condition of the Steam Generators The steam generators at Millstone 2 are Combustion Engineering Model 6F with 8,519 tubes per steam generator.
To maintain operability of the steam generators the licensee conducts an inservice inspection-program as prescribed in technical specifications 4.4.5.1.2 and 4.4.5.1.3 during refueling outages.
The inservice inspection results indicate the number of tubes that are defective and required to be taken out-of-service.
The limiting size for a tube defect that would require plugging is equal to 40*4 of the nominal wall thickness.
The licensee implemented two methods to repair defective tubes, plugging and sleeving.
In May, 1983, the licensee installed tube sleeves for ' pit' degradation between the tube sheet and the first support plate.
Previous to and after the tube sleeve installation, defective tubes were removed from service with tube plugs. The current number of tube sleeves and plugs in the steam generators are listed below:
- of Plugs No I 2370 3482 No. 2 2281 2362 In previous inspection reports (50-336/89-05, 50-336/89-08 and 50-336/89-17) the inspector documented licensee actions in response to failures of mechanical tube plugs, and their response to NRC Bulletin 89-01 " Failure of Westinghouse Steam Generator Tube Mech-anical Plugs." Those actions included development and installation of a plug-in plug (PIP) for all susceptible tube plugs. The current number of PIPS installed in the steam generators is:
Steam Generator Number of PIPS No. 1 247 No. 2 199 Both the PIP and the sleeve installations are designed as ' leak
!imiting' devices for primary to secondary leakage.
Plant Design Change Request (PDCR) 2-72-83 specifies.that the design leakage from 3000 sleeves will not exceed 0.02 gpm, and PDCR 2-088-89 states that the average PIP installation will not exceed 0.003 gpm with a maximum leak of 0.01 gpm.
The licensee concluded both designs were not an unreviewed safety question as defined by 10CFR50.59. Resident and specialist inspection of licensee actions for NRC bulletin 89-01 found the leak limiting devices for suspect tubes adequate.
PDCR
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5 2-72-83 and 2-088-89 safety evaluations also considered the consequences of postulated events including the failure of the sleeve or PIP, in conjunction with postulated SG tube rupture events, and primary and/or secondary depressurization events. NRC review of the licensee's safety evaluation for the plug and sleeve designs identified no inadequacies. The steam generators at Millstone 2 are designed with leak limitirg devices such that primary to secondary leakage pathways potentially exist, but with design margins to ensure safe operation of the facility.
Off-Site Oose Assessment The inspector reviewed the licensee's assessment of off-site dose due to releases from the steam jet air ejector. The inspector referred to Technical Specification 6.15, the Offsite Dose Calculation Manual (ODCM); 10 CFR 20.106; and 10 CFR 20.106(7)(d).
The licensee's calculation used conservative assumptions for total SJAE activity of 3X10(-4) microcuries (uci)/ cubic centimeter (cc)
and a ' spike' lasting 1 minute per day. The calculations showed that the SJAE activity observed as of April 7 would result in 0.0003% of the allowable off-site dose release rate and an annual dose of 4x10(-6) millirad / year exposure to a hypothetical person standing continuously at the site boundary.
The limits at the site boundary for a gamma dose is 5.0 man-rem / quarter, and 10 man-rem / year.
The allowable release rate at Millstone station would result in 1.3 millirad / year.
In conclusion, for the SJAE activity and release rates observed during this inspection period, the release of radia-l tion to the public is significantly less than regulatory lirrits and is negligible.
l Procedural Requirements for Leakage The inspector reviewed licensee procedures for a primary to secondary leak to assess the adequacy of required actions at the l
facility.
Procedures reviewed are listed below:
- OP 2260 " Millstone 2 Emergency Operating Procedure (EOP) Users Guideline," Rev. 1
- AOP 2569 "SG Tube Leak" Rev. 3 l
- OPS Form 2569-1 " Estimate of Leakage"
- E0P 2534 "SG Tube Rupture"
- SP 2833 " Secondary Analysis for Total Gamma Activity"
- CP 28064 " Calculating Primary to Secondary Leak Rates"
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Inspector review determined that licensee procedures were adequate j
to direct operator responses to alarms to diagnose a primary to
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secondary leak, to increase radiochemistry sampling in response to indications of increases in leakage rate, and to evaluate radio-
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chemistry data to estimate the magnitude of the leak rate.
In AOP i-2569, the decision to continue operation with a known leakage resides
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with plant management, and if technical specifications for leakage or
SG activity are exceeded, a shutdown of the plant is required by procedure and the applicable technical specification action. state-ment..The inspector noted during discussions with unit management and shif t operating personnel that the licensee would take actions to shutdown the plant prior to exceeding the technical specification limits.
Control room alarms that provide indication of a primary to secondary leakage are:
" Main Steam Line High Radiation / Instrument Failure,"
process monitor radiation high and local alarms at the RM-4262,
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RM-5099, and the N-16 monitors. Operator actions in response to
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indications of increased steam generator leakage are described in procedure OP 2316A and plant operations night order 7/14/89-1. Once the steam jet air ejector radiation monitor (RM-5099) alarms or the steam generator blowdown monitor (RM-4299) alarms and isolates the steam generator blowdown, control room operators are required to request the chemistry department to sample'either the steam jet *
ejector or blowdown, as applicable.
The operator should then docu-ment the time of the alarm in control room logs, the automatic actions that occurred, the chemistry sample results, and the corres-
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ponding readings from the N16 radiation monitors.
Inspector obser-vations of control room activities noted that the operators performed the actions as directed by the procedures.
The inspector noted a heightened awareness to leak rate trends in control room personnel and at the licensee daily work planning meet-ings. Utility management actions in response to current leakage rates included (1) installation and use of an additional air ejector monitor; (2) confirmation of leakage rates by tritium analysis; (3)
the use noble gas half-live ratios to determine an approximate time delay in the leak path between primary and secondary systems; and, (4) evaluation of air ejector radiation " spikes".
Conclusion
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At the end of the inspection period, the licensee has a confirmed
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primary-to-secondary leak. The leakage values are generally a small
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fraction of the technical specification limits. While some uncer-tainty existed regarding the exact leak path, the major indications from the transport mechanism was that an indirect leak path existed, rather than a more significant degradation in an inservice tube.
Significant licensee management and control room operator awareness of current leakage values and trends were noted.
The licensee's primary to secondary leakage rate monitoring program is considered good based on management initiatives to quantify low level leakage, b
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the implementation of additional monitoring capabilities not required in TS, and additional sampling evaluations by the chemistry department (i.e., tritium predictive analysis).
3.5 Cable Raceway Separation On March 21, the inspector toured the facility cable spreading room to review the configuration relative to commitments for cable raceway separation between redundant safety-related trains. The reference used was the Final Safety Analysis Report (FSAR) section 8.7.3.1, which states that separation of redundant cables is accomplished by spatial separation of cable trays. The spatial separation required is not less than four feet vertically and eighteen inches horizontally to guard against damage from external fire, missile, or other hazards.
Six specific locations were identif;ed without the specified minimum spatial separation as documented in the FSAR.
The cable raceway identifications are as follows: Z15RB10/Z24LB70, Z15RM10/Z25LD10, Z15RD10/Z24LB60, 225LG10/Z15RE10, 215RH10/Z24LB35, and 215RK20/Z24LB25.
The worst case cable-raceway separation identified was three inches vertically.
In response to the inspector's observations, the licensee initiated
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Plant Incident Report (PIR) 90-16 on March 21 to document identified discrepancies in cable raceway separation, and initiate compensatory fire protection patrols.
TS 3.7.10.a.1 requires all fire rated I
assemblies (i.e. cable tray enclosures) separating safety-related fire areas or portions of redundant systems for safe shutdown within designated fire areas to be operable. According to NUSCo Drawing 25203-34077 Sheet 14 " Appendix R Electrical Tray and Conduit' Plan, Cable Vault," the six identified cable raceways were associated with redundant safety trains.
The licensee's immediate corrective actions included:
establishing compensatory fire protection patrols; review of plant inspections and design changes, and in plant walkdowns on March 22 and 23 to further review cable spatial separations with respect to FSAR 8.7.3.1.
The licensee's review concluded that post-construction I
changes per plant design change record (PDCR) 2-76-81 incorporated
the use of marinite boards to compensate for reduced cable raceway
separation in the cable spreading room. Additionally, the licensee walkdowns identified one zone in the auxiliary building near the spent fuel pool skimmer equipment that contained a cable raceway cross-over that did not comply with the FSAR commitments.
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i On November 19, 1980, the NRC published a revised Section 10 CFR50.48 and a new Appendix R to 10CFR50 regarding fire protection features at nuclear power plants.
The revised section 50.48 and Appendix R became effective on February 17, 1981.
10CFR50 Appendix R contains subsections A thru 0, each of which specifies requirements for a particular aspect of fire protection features. The inspector
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reviewed the exemption requests for the Fire Area 24, and the Appendix R Fire Area R-1, Cable spreading room.
On April 15, 1986, the NRC granted exemption requirements from 10CFR50 Appendix R Section III.G. for the cable spreading room. The exemption allowed redundant shutdown related cables. hat were not separated by a 1-hour fire-rated barrier or more than 20 feet with no combustibu terials. The bases for the exemption were that the cables in thi.
c,e spreading room were qualified to Institute of
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Electrical ana Electronic Engineers Standard (IEEE)-383 or coated with fire retardant material.
IEEE-383 qualified cables are made such that a postulated fire within the cable does not generate enough heat to ignite a fire affecting a redundant cable tray. The
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cable room fire detectors were developed under project assignment
!77-597 to detect a fire before damage to two trains could occur; and I
an in-tray suppression system was installed by the licensee under project activity 77-604. The exemption request concluded that the spatial separation between redundant trains achieves sufficient
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passive fire protection to provide reasonable assurance that one i
shutdown train would remain free of damage.
In conclusion, based on
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the 10CFR50 Appendix R Section III.G exemption, the present
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condition of cable separation in the cable spreading room is accept-able, i
The inspector reviewed the spatial separation relative to postulated
internal electrical fault propagation from one safety train cable to a redundant cable. The inspector's review considered NRC Regulatory
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Guide.1.75 " Physical Independence of Electrical Systems." Millstone
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2 is not committed to NRC Regulatory Guide 1.75; however, the
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licensee is committed to section 4 of IEEE 279-1971 and section 4 and 5 of IEEE Standard 308, 1971 for separation of conductors and I
wireways.
The separation of cable trays for electrical fault propagation is acceptable since the issue was addressed by the licensee in the original plant design basis construction through Bechtel specification 25203-34001, and qualification of cables per IEEE-383.
The inspector has no further questions in this regard.
To this end, the as-found configuration of cable tray separation in the cable spreading room was acceptable.
The licensee's follow-up plant walkdowns identified one specific installation where commitments to FSAR 8.7.3.1 were not acnieved.
The Fire Hazards Analysis for the spent fuel pool skimmer area identified a fire loading of 23 minutes with a detection and automatic suppression system.
The area was identified by the licensee and is a regularly travelled access path for operators, in addition the cables were
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IEEE-383' qualified. The update of the-FSAR 8.7.3 to reflect 10CFR50 Appendix R Section-III.G. exemption-for the specific fire area is an unresolved item (50-336/90-06-01).
3.6 Previously Identified Items
~. 3. 6.1 (Closed) Unresolved Item 89-22-02:
Evaluation of Fire Consequences With An Open Fire Barrier This item was open pending licensee evaluation of the conse-quences of a postulated fire without maintaining sufficient fire barrier seal between the cable vault and west DC switchgear room. The unsealed fire barrier was an open two-inch conduit
passing between fire areas A21 (west DC switchgear room), and l
c A24 (cable spreading room), and extending approximately twelve
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feet into the cable spreading room.
Licensee internal document (GMB-89-271), dated November 9,1989, assessed the consequences of the open barrier in question. The document specifically addressed fire initiation in each fire area, involvement of operable suppression. systems, and the poten-tial for spread of the postulated fire into the unaffected zone via the open barrier.
The matter was' reviewed by NUSCO fire protection engineering.
The licensee's conclusion was that the open barrier did not pose a significant_ hazard that would affect the capability to achieve safe shutdown.
Inspector review and-1 discussions with licensee personnel concluded that the open fire q-barrier posed no adverse threat in achieving safe plant shutdown
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This item is, closed, j
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4.0 Radiological Controls (IP 71707/92701)
. m '.. of Radiological Areas
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4.1 Posting ant
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During ple-totc s, contaminated, high airborne radiation, and high radiation.reas were * 'iewed with respect to boundary identification, locking r,; a ament ; e appropriate control points.
No inadequacies
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were note (
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.wd, w.lon exposure for March was 0.734 man-rem, which was v c a...w.
,o the licensee monthly operational goal of 5.5 man-rei..
" e aagement used the daily planning meetings to
-provide good discussion of and heightened awareness of actions to-i keep' radiation exposure as-low-as-reasonably-achievable (ALARA) for routine activities.
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5.0 Maintenance / Surveillance (IP 62703/61726/92702)
5.1 Observation of Maintenance Activities l
The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with regulations,.
use of_ administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel prote: tion, and equipment alignment and retest. The following activity was included:
-- AWO Mr@)MM " Fire Pump Packing Adjustment"
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No inadequec hs were identifid 5.2 Observation of Surveillance Activittes The inspector observed portions and review of completed surveillance tests to assess performance in accordance with approved procedures
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and Limiting Conditions of Operation, removal and restoration of equipment, and deficiency review and resolution. The following tests were reviewed:
-- SP-2404BC " Main Steam Line Radiation Monitor Function Test"
-- SP-2401F " Reactor Protection System, High-Power Trip Functional
. Test"
-- SP-2404AV " Steam Jet Air Ejector Functional Test and Calibration" No inadequacies were noted.
5.2.1 Review of SP 2404AW for Calibration of Radiation Monitor P.M 6038 An employee presented a technical question to the inspector on March 6, 1990 regarding surveillance procedure SP 2404AW, Liquid Process Radiation Monitor (RM 6038) Calibration, for calibrating the reactor building closed cooling water (RBCCW)
radiation monitor. The question was whether the procedure was adequate if a " reproducible geometry" was not assured when ireasuring background radiation levels for the background correction. Inspection of this matter included review of SP 2404AW and completed test results for the surveillance.
Additional references and further details of the inspector's review are provided in Attachment B.
The inspector noted that the test method in the current version of SP 2404AW (revision 1) does not use a blank cartridge filled with deionized water to obtain the " background" count rate for the detector.
Instead, per step 7.1, the normal sample canister is removed from the lead-shielded housing, the detector is re-installed into its location inside the housing,
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Test data-reviewed by
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the inspector. showed that the background readings varied from about 13 cpm to as high as 110 cpm, depending on the presence and varying strength of radiation fields from other sources in the area of the radiation monitor.
The inspector noted that the purpose of the background check is to correct the detector reading for radiation entering the detector from all sources other than from the radionuclide concentration in the process fluid. Since the " background" radiation comes from all directions in a "4 pi" geometry, and
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not just from the geometry presented by the sample process
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canister, it is less important to maintain a source-detector geometry for the background determination than it is when the detector response is determined for exposura to the three.
calibration sources. A fixed detector-source geometry is important when the calibration sources are used to assure good correlation between secondary and primary calibrations on the monitor, when radioactivity in the process canister (effluent stream) is present as the source of radiation.
Inspector calculations using the very conservative assumption that the background determinations are off by 100*4, showed that the error would be most significant for the isotopic calibration on the lowest end of the range, and negligible at the upper two
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calibration points. While use of a blank cartridge. filled with deionized water would make the background measurement more
. accurate, the error introduced into the calibration is insignificant.
Based on the above, the inspector concluded that the present calibration method for RM 6038 was adequate and would result in an acceptable calibration of the monitor that would meet regulatory requirements.
Notwithstanding this conclusion, the practice of using a blank cartridge for background determinations, a licensee initiative to improve the calibration technique, would be a desirable enhancement.
5.3 Previously Identified items 5.3.1 (Closed) Unresolved Item 88-24-01:
Incorporation of Material Equipment Parts List Quality Assurance Indicators into the Production Management Maintenance System This item was initiated based on. inspector review of fire protection instrument quality assurance (QA) indicators in the production management maintenance system (PMMS) system.
The review indicated discrepancies between material equipment parts list (MEPL) QA indicators and those incorporated in the automated work order PMMS system.
At the time of the prior inspection, no loss of work controls were noted; however, the open item concerned the potential implications of a programmatic
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breakdown in transcription of MEPL QA indicators into the automated PMMS system.
In late 1988, the licensee acquired contractor services to review
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all instruments in the plant initially evaluated and listed in the MEPL with an undetermined status for a QA indicator. The review lasted approximately ten months. Independent evaluations are performed in the corporate engineering department. At the time of this inspection, 506 MEPL evaluations were completed with initial and independent reviews completed.
l Approximately 44 evaluations still require an independent y
evaluation. All work activity associated with the instruments-
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having the incomplete evaluation is controlled as a QA safety
classification, until the MEPL evaluations have been verified.-
i The inspector randomly selected fourteen in plant instruments
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and compared the MEPL evaluation results and the PMMS work l
order classifications. The listing below shows the instruments i
reviewed:
INCORPORATION INSTRUMENT MEPL EVAL INTO PMMS TS-8842 Vital AC Cooling CD*756 Yes
Fan
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PD1-8155 Temperature Switch CD*725 Yes Fan Prefitter Differential Pressure i
FI-212 Charging Header CD*572 Yes to Regenerative HX Flow l
RM-8123 MP2 Stack Monitor CD*802 Yes
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LC-6000 RBCCW Surge Tank CD*516 Yes
LY-110Q Current to Pressure CD*591 Yes Converter l
T-112CA Rx Coolant Loop CD*616 Yes 1A Cold Leg Temp
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P-100X Pzr Pressure CD*682 Yes
Control Loop
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T-3034 LPSI SDHX Outlet CD*666 Yes PI-7303 Fire Pump Discharge CD*1017 Yes Pressure TS-7822 Instrument Air CD*614 Yes
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Dryer Heater Failure HV-8001
'A' Control Room CD*746 Yes Exhaust Fan Dryer LIC-8242 Containment Sump CD*851 Yes Level PI-8729 Starting Air CD*873 Yes Pressure for 'B'
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Based on the above review,_the inspector identified no inadequacies in the licensee's program to update the PMMS QA indicators to reflect MEPL classifications, and no programmatic breakdowns were noted.
The inspector considers this item closed.
6.0 Engineering / Technical Support (IP 92702)
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6.1 Pressurizer Cooldown Rate Discrepancy On April 10 at approximately 2:15 p.m., the licensee reported to the NRC in accordance with 10CFR 50.72(b)(1)(1)(i) a plant condition that was outside the original design basis.
The technical specification (TS) requirements were not consistent with the reactor coolant system pressurizer fatigue stress analysis.
Specifically, JS 3.4.9.2.b limits pressurizer cooldown rate to a maximum value of 200 degrees per hour. The basis of the TS cooldown limitation is to assure the pressurizer is operated within the design criteria
assumed for the fatigue analysis performed in accordance with
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American Society of Mechanical Engineer (ASME) code requirements.
The nuclear steam supply system supplier recently provided a fatigue analysis that concludes the cooldown rate should be limited to 100 degrees F/hr, between 635-355 degrees F and 200 degrees F/hr _between 355 degrees F and ambient.
Thus the TS 3.4.9.2.b cooldown limits were not supported by the fatigue analysis and were thus
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non-conservative.
Licensee corrective action for the non-conservative TS pressurizer cooldown rates was to implement administrative controls to not exceed 100 degrees F per hour during a plant cooldown.
The administrative controls were promulgated in the plant operations night orders.
-The licensee's determination that the issue was reportable to the NRC was based on a review of all historical cooldown rate data to determine if the fatigue analysis results had been exceeded.
Licensee review identified five instances where cooldown rates exceeded the fatigue analysis.
All five instances involved cooldown evolutions with initial reactor coolant system temperature above 355 degrees F and that ended below 355 degrees F, i.e.,
in the cooldown rate transition area where the limiting rate changes from 100 to 200 degrees F per hour.
The licensee determined continued plant operation is acceptable in the short term in spite of the identified instances of exceeding the fatigue analysis for pressurizer cooldown.
This conclusion is based on the fact that the Millstone 2 pressurizer is designed with the same sizing calculations for wall thickness, material, and radius as other Combustion Engineering plants, and Combustion Engineering evaluation of pressurizer material properties for other similar plants shows an acceptable fatigue analysis for a 200 degrees F/hr
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cooldown rate for all of ASME Section III design criteria.
The
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inspector had no further questions regarding the acceptability of'
operation.
The inspector questioned the communication between corporate and site engineering and the timeliness of the reportability determina-tion. A chronology of events concerning the pressurizer cooldown rate at Millstone 2 is given below:
Date Event February 23, 1990 CE0G-90-159:
A nuclear facility requested a f
conference call to discuss the cooldown rate
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differences between the TS basis and the fatigue i
analysis, March 6, 1990 CE0G-90-190:
Representatives of six utilities i
(including Northeast Utilities) decide issue is i
not reportable under 10CFR 50.72 unless fatigue
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analysis results were actually exceeded.
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March 19, 1990 CEOG-90-203:
Comoustion Engineering provides
cooldown rate fatigue analysis values for l
affected facilities.
March 29, 1990 Resident discusses issue with site engineering supervisor.
The licensee initiates plant incident report 2-90-18.
I April 2, 1990 The licensee initiates reportability-evaluation
per procedure NE0 2.25 under 10CFR
50.73(a)(2)(1)(B)
i-April 4, 1990 Millstone 2 Unit Director notified per procedure
'l NE0 2.25.
April 10, 1990 Licensee reported pressurizer cooldown rate I
outside original design basis per 10CFR l
50.72(b)(1)(1)(1).
A weakness was noted in the normally aggressive communication and corrective action between corporate and site engineering, in that the resolution of an issue that affected the facilities operating basis was delayed for approximately one month. The unit director acknowledged the inspector's concern on April 27.
The inspector considers this item unresolved pending licensee action to reconcile the technical specification cooldown rate limiting con-dition for operation with the updated fatigue analysis. (UNR 50-336/
90-06-02).
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6.2 Reactor Fuel Performance
'On April 11, the inspector discussed reactor mid-cycle fuel
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performance with the licensee's reactor engineering personnel. The
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review included fuel vendor evaluation of reactor coolant system (RCS) radio-chemistry results, RCS gaseous, flash gas, and iodine activity during cycle 10 operation, TS requirements and basis, and reporting criteria.
The fuel vendors (Westinghouse, and Advanced Nuclear Fuels) have provided two estimates of failed fuel based on radiochemistry results: 3-5 fuel pin failures based on RCS radio iodine concentrations, and 6-8 fuel pin failures based on RCS Xenon gas concentrations.
The dose equivalent iodine measurement for conditions on April 12 was 0.0152 microcuries (uci)/ milliliter (ml) and total degassed radioactivity was 0.386 uci/ml.
The values have trended upward from initial values of 0.009 and 0.19 uci/ml, respectively, at the beginning of cycle 10 operation (May 3, 1989).
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The technical specifications require specific activity of the primary coolant be limited to less than or equal to 1.0 uci/ml of dose equivalent iodine.
The basis for the limits on specific activity assure that the resulting 2-hour doses at the site boundary will not exceed a small fraction of 10 CFR 100 limits following a postulated steam generator tube rupture in conjunction with an assumed steady state primary to secondary leak of 1.0 gallon per minute (gpm). The inspector concluded that present RCS radiochemistry-parameters are well within the technical specification limits and-plant operation with continued monitoring is acceptable,
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The current licensee plans for possible future corrective actions based on fuel performance include either; reconstitution of fuel, non-destructive examinations of fuel assemblies, or routine refuel activities. The inspector will follow licensee corrective actions on fuel performance in future inspections.
7.0 Security (IP 71707)
Selected aspects of site security were verified to be proper during inspection tours, ' including site access controls, personnel searches,
_l personnel monitoring, placement of physical barriers, compensatory measures, guard force staffing, and response to alarms and degraded conditions.
Two security events per 10 CFR 73.71(c) were reported to the NRC during the inspection period. One event involved a breech between the owner-controlled area and the protected area, and the other event
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involved the discovery of contraband within the protected area.
Licensee corrective action and investigations were well-developed, The inspector
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reviewed the events and consulted regional specialists and had no further questions.
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8.0 Safety Assessment / Quality Verification (IP 30703/40500/90712/92702/92700)
8.1 Committee Activities The inspector attended meetings 2-90-23, 2-90-28, and 2-90-29 of the Plant Operations Review Committee (PORC) on March 12, March 21, and March 22.
The inspector noted by observation that committee administrative requirements were met for the meetings, and that the committees discharged their functions in accordance with regulatory requirements.
The inspector observed a thorough discussion of matters before the PORC and a good regard for safety in the issues under consideration by the committee.
No inadequacies were identified.
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8.2 Nuclear Review Board The inspector attended a portion of the Millstone 2 Nuclear Review Board (NRB) meeting 2-90-04 on March 20.
The technical specification requirements were satisfied for composition, meeting frequency, quorum, and topics for review.
The NRB devoted most of the meeting to a review of the corporate engineering majority view regarding the steam generator safety assessment, and the inter-departmental disagreements with the assessment conclusion. The assessment was documented to the NRC on February 7,1990 and was an input to discussions in a meeting-between Northeast Utilities and the NRC on February 22.
The discussions before the NRB included justification for the safety assessment, engineering judgement disagreements related to the analysis, consideration of a need for a unreviewed safety question-(i.e. 10 CFR 50.59) and justification for use of-boric acid treatment in the secondary plant. The boaro-deferred a conclusion regarding the engineering assessment pending receipt of further information.
The NRB devoted the remainder of the meeting to plant incident report (PIR) tracking and control, and various proposed technical specification and proposed amendment requests. The inspector noted each member was provided a meeting briefing package containing copies of the documents scheduled for review. The board members provided informed assessment and evaluation. The inspector has no further question U i.
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8.3 Periodic Reports Upon receipt, periodic reports submitted pursuant to technical specifications were reviewede This review verified-that the reported information was valid and included the required NRC data.
The inspector also ascertained whether any reported information should be classified as an abnormal occurrence.
The following
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reports were reviewed:
Monthly Operating Report - February,1990 l
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Monthly Operating Report - March, 1990
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Special Report, per Technical Specification 3.3.3.1 Table
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3.3-6, Action 17 on High Range Stack Radiation Monitor No inadequacies were noted.
9.0 Management Meetings Periodic meetings were held with station management to discuss inspection
findings during the inspection period. A summary of findings was also discussed at the conclusion of the inspection.
No proprietary information was covered within the scope of the inspection.
No written material was
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given to the licensee during the inspection period.
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9.1 Corrections to Previous Inspection Reports l
Inspection Report 50-336/89-22 Section 5.5, Paragraph 8, identified regulatory requirements for-the operability of fire seal barriers for internal conduits. The requirements referenced were 10 CFR 50, Appendix R, Section II. A and Section II.B.4, and technical specification 3/4.7.10.
Licensee letter " Fire Barrier Penetration Seal Program Upgrade" dated September 28, 1988, identified that Millstone 2 was licensed to the May, 1976 version of Appendix A to Branch Technical Position (BTP) 9.5-1, which requires sealing around components penetrating the designated fire barrier boundaries (i.e. piping, cable trays, conduits) as prescribed in technical specification 3/4.7.10.
In
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subsequent revisions to Appendix A to BTP 9.5-1, and then issuance of Appendix R to 10 CFR 50, sealing inside of conduits penetrating fire boundaries became a regulatory requirement for those plant
. licensed to later revisions. By letter _ dated March 31, 1989, the NRC identified the above licensee's position as an accurate representation of the NRC position.
, Inspection Report 50-336/89-05 (A.19.03)
Section 11, Page 24, Paragraph 5 stated "the inspector reviewed the setpoint change issue with the operations and I&C departments, who
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stated that the area radiation monitor (ARM) (RM-7890/7891) does not undergo a setpoint change for refueling operations."
The inspector was informed on August 8, 1989 that the containment ARMS undergo a setpoint change as required in health physics procedure 904/2904/39040 figure 6.5.
The licensee clarified procedure requirements regarding setpoint changes by additions to operations procedure 0P2383B, the pre / post refueling checklist, and to the I&C calibration procedure.
9.2 Concerns Referred to the Licensee for Resolution On February 16, March 5 and March 6, the inspector presented various concerns from employees at Millstone station to licensee management for resolution. The issues were entered into the licensee's nuclear concerns program for tracking. Attachment A to this inspection report lists the issue number, the date received by the NRC and the concern or issue.
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No unsafe conditions were identified.
Licensee actions to respond to employee identified concerns will ba reviewed on subsequent routine inspections.
9.3 Concerns Reviewed by the Inspector ( A.45.011 The following specific concern raised by an employee on March 6, 1990 was reviewed during this inspection. The issue involved the technical adequacy of surveillance procedure SP 2404AW, Liquid Process Radiation Monitor (RM 6038) Calibratien, for calibrating the reactor building closed cooling water (RBCCW) radiation menitor.
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The question was whether the procedure was adequate if a " reproducible geometry" was not assured when measuring background radiation levels for the background correction. The employee stated that the procedure approach was contrary to information recently provided during a technical training course on radiation detector calibrations.
The question was previously presented to I&C supervision, who responded by memo dated February 23, 1990 that the existing procedure method was acceptable as is, and an intended change to the procedure to better account for geometry in the background determination was judged to be an enhancement.
Inspection of this matter included inspector review of the response from the I&C supervisor; review of the pertinent training material; and, review of SP 2404AW, including completed test results.
Additional references and further details of the inspector's review are provided in Section 5.2.1 above and in Attachment B.
Inspector review found that calibrations using the test method of SP 2404AW would result in an acceptable calibration that meets regulatory requirement,
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ATTACHMENT A Allegations Presented to the Licensee Management for Disposition Number Date Alleged Concern A.33.01 February 16 The wide range NI calibration deficiency 1990-A.33.02 February'16 The pressurizer surge tank level transmitter 1990 cable noise.
B.'14.01 February 16 Conflict between two administrative procedures 1990 on description of job supervisor.
B.15.01 February 16 Tagging for transformer draw and main generator 1990 link, 2720G3 were improper.
B.17.02 February 16 Test of fire detectors in TG, procedural-
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1990-inadequacies.
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8.17.04 February 16 Improper access controls for EEQ material 1990 i
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o ATTACHMENT B REVIEW 0F TECHNICAL CONCERNS ON CALIBRATION OF MONITOR RM 6038 A technical question by a Millstone 2 employee was presented to the inspector on March 6, 1990 regarding surveillance procedure SP 2404AW, Liquid Process Radiation Monitor (RM 6038) Calibration, for calibrating the reactor building closed cooling water (RBCCW) radiation monitor.
The question was whether the procedure was adequate if a " reproducible geometry" was not assured when meas-uring background radiation levels for the background correction.
The employee stated that the procedure approach was contrary to information recently provided during a technical training course on radiation detector calibrations.
The question was previously presented to I&C supervision, who responded by memo dated February 23, 1990 that the existing procedure method was acceptable as is, and an intended change to the procedure to better account for geometry in the background determination was judged to be an enhancement.
Inspection of this matter included inspector review of the response from the I&C supervisor; review of the pertinent training material; review of SP 2404AW;
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discussions with engineering personnel in the licensee's radiological assess-ment branch; review of the radiological assessment branch reports " Radiation Monitor Review - Phase 1 and 2"; review of completed surveillance test results
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for SP 2404AW dated 8/1/89, 11/1/89, and 2/1/90; and, consideration of the cri-teria and guidance in Regulatory Guide 4.15 and ANSI Standard N13.10, "Specifi-cation and Performance of On-Site Instrumentation for Continuously Monitoring Radioactive Effluents."
RM 6038 monitors the RBCCW system and will alarm at preset limits to alert operators to the presence of radioactivity in the RBCCW system.
RBCCW is a closed system and is normally radioactively clean.
The purpose of RM 6038 is to monitor for gamma activity that would be indicative of heat exchanger leak-age from systems cooled by RBCCW (e.g. the shutdown cooling heat exchangers).
The monitor is required to be operable by technical specification (TS) 3.3.3.9 to act 'as the final ef fluent monitor for the service water system. Detectable activity in service water would occur only if RBCCW activity was high and an additional leak occurred from the RBCCW system to the service water system.
The alarm setpoints for RM 6038 are determined in accordance with the method-ologies in the offsite dose calculation manual, which are set to assure the 10 CFR 20 limits on radioactivity in liquid effluents are not exceeded.
RM 6038 is a gamma scintillator that uses a 2 inch by 2 inch sodium iodide de-tector, that is highly sensitive to radiation.
Based on a typical conversion factor of about 6 X 10-9 uci/mi per counts (in cpm) of monitor response, the monitor measures radioactivity over a range of I cpm to 10+6 cpm, which cor-responds to an isotopic concentration in the range of 6 X 10-8 uci/ml. The secondary calibration technique used in SP 2404AW is to expose to the detector three concentrations of solutions containing Cs-137 in a liquid sample con-tainer of equivalent geometry to the process chamber. The detector response for each source is plotted to obtain an average calibration facto '
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The inspector noted that the test method in the current version of SP 2404AW (revision 1) does not use a blank cartridge filled with deionized water to ob-tain the " background" count rate for the detector.
Instead, per step 7.1, the normal sample canister is removed from the lead-shielded housing, the detector is re-installed into its location inside the' housing, and the detector readout is recorded. The data for all tests reviewed in the period noted above showed that the recorded background readings varied from about 13 cpm to as high as 110 cpm, depending on the presence and varying strength of radiation fields from other sources in the area of the radiation monitor.
The inspector noted that the purpose of the background check is to correct the detector reading for radiation entering the detector from all sources other than from the radionuclide concentration in the process fluid.
Since the
" background" radiation comes from all directions in a "4 pi" geometry, and not just from the geometry presented by the sample process canister, it is less important to maintain a source-detector geometry for the background determina-tion than it is when the detector response is determined for exposure to the three calibration sources. A fixed detector-source geometry is important when the calibration sources are used to assure good correlation between secondary and primary calibrations on the monitor, when radioactivity in the process canister (effluent stream) is present as the source of radiation.
In addition to the intuitive evaluation given above, the inspector assessed the maximum error introduced by the test method in the existing procedure.
In ad-dition to the existing lead and steel housing that shield the detector, the presence of a blank process canister filled with water would further shield the
. detector from background sources, and would thus make the measured " background" less (a smaller number).
Recording the background without a blank canister introduces a systematic error in the calibration whose effect is to always make the background correction factor slightly greater than it might otherwise be.
Inspector calculations using the very conservative assumption that the back-ground determinations are off by 100?;, showed that the error would be most sig-nificant for the isotopic calibration on the lowest end of the range, and neg-ligible at the upper two calibration points.
For the 2/1/90 test, the monitor was source calibrated at 7000, 2 X 10+5 and 7 X 10+5 counts per minute.
The error introduced by the background correction ranged from 0.18?4 to 1.4?; on the lowest test decade. While use of a blank cartridge filled with deionized water would make the background measurement more accurate, the error introduced into the calibration is insignificant.
Inspector review of the monitor performance for the calibration completed on 2/1/90 noted good detector linearity throughout its range, in that all ratios of the source corrected counts per minute to the decay corrected isotopic source strengths were within +/-20 percent.
If the error noted above was accounted for in the decay corrected source strength to detector cpm ratios, the detector linearity would still be acceptabl :6?
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The inspector noted further thatiprocedure acceptance criteria was met, or dis-
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crepancies were.noted and'dispositioned through instrument calibration reports (ICRs).- Based on-the above, the inspector concluded that.the present calibra-
. tion' method-for RM 6038 was adequate and would result in an acceptable calibra-tion' of the monitor.that would meet regulatory requirements.
Notwithstanding
this conclusion, the practice of using a blank cartridge for background deter-minations was recommended by the licensee radiological assessment branch in the-
' Millstone 2 radiation monitor manual : issued on-April 1,1990.- Licensee initi-
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.atives to improve the calibration: techniques by adding the noted. enhancements are. desirable.;
Based on the above, the inspector identified no inadequacies in the.supervi-
'sor's L Februa ry 23,1990 memorandum.
The inspector had no further questions -in
'this area.
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