ML20155F851
| ML20155F851 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/29/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20155F843 | List: |
| References | |
| 50-213-98-04, 50-213-98-4, NUDOCS 9811060135 | |
| Download: ML20155F851 (59) | |
See also: IR 05000213/1998004
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U.S. NUCLEAR REGULATORY COMMISSION
i- REGION I
Docket No.: 50-213
License No.: ' DPR 61
l. Report No.: 50 213/98-04
Licensee: Connecticut Yankee Atomic Power Company
P. O. Box 270
Hartford, CT 06141-0270
Facility: Haddam Neck Station
Location: Haddam, Connecticut
Dates: . July 20 - September 11,1998
Inspectors: Ronald Burrows, Project Manager, NRR
Thomas Fredrichs, Project Manager, NRR
Marie Miller, Senior Health Physicist
Lonny Eckert, Radiation Specialist
Christopher Wolch, Reactor Engineer
John Wray, Decommissioning Health Physicist
Joseph Nick, Decommissioning Health Physicist
William J. Raymond, Senior Resident inspector (Lead inspector)
Approved by: Ronald Bellamy, Chief, Decommissioning and Laboratory Branch
Division of Nuclear Materials Safety
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9811060135 981029
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EXECUTIVE SUMMARY
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Haddam Neck Station
NRC Inspection Report No. 50-213/98-04
This special inspection included aspects of licensee preparations, planning and
implementation of the reactor coolant system (RCS) chemical decontamination. The report
covers a two-month period of inspection by resident, regional and NRC Headquarters
personnel, and includes reviews and assessments of operations, engineering, maintenance,
and plant support activities, and management effectiveness.
Decsw.m.lee:sn r,s Onorations and Saaaa t Activities:
The RCS chemical decontamination accomplished substantial dose reduction in reactor
piping, but was hampered by several events that challenged plant personnel and the safe
control of the radiological source term. Licensee preparations for and the conduct of the
RCS decontamination were generally adequate, but many weaknesses were evident in -
several events and equipment problems. The events included two major leaks of
decontamination fluid, the loss of control of domineralizer resins resulting in elevated
radiation doses in plant piping, and the loss of control of a 5-ton floor block due to
improper rigging. Plant staff failed to exhibit conservative decision-making in several
instances, which contributed to some events. _ The licensee failed to adequately assess the
material condition of the plant, particularly in the letdown and purification systems.
Numerous demineralizer system failures hampered the smooth conduct of the
decontamination. Several weaknesses were noted in the support for the decontamination,
such as ln the evaluations for the operation and attachments for the letdown booster
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pump; the operation of the C domineralizer and the replacement of the post-filter retention
element; and, the review of floor block rigging. Apparent violations of Technical
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Specification 6.8.1 related to inadequate procedures included: 1) failure to perform I
adequate leak checks of the RCS decontamination boundary; 2) failure to adequately
address potential transient conditions in the letdown system equipment; (3) failure to
assure that activated resins would not pass through the lower retention element of the
purification system under high flow conditions during the RCS decontamination; and (4)
failure to provide procedural guidance and worker training on how to rig a pipe trench floor
block that contributed to the loss of control of a five ton block.
The plant staff responded properly to events to mitigate adverse radiological conditions,
and took appropriate actions to address degraded conditions. Following the identification
of degraded piping, repairs were acceptable and the pressure tests provided assurance of
the leak tightness. Engineered clamps used on temporary decontamination piping were
acceptable. Repairs of permanent piping were completed per ANSI B31.1. Following the
July 27 leak, engineering evaluations and thermal-hydraulic modeling performed for the
static and expected operating conditions were adequate to ensure the systems were
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capable of supporting the anticipated operating parameters during the remainder of the '
decontamination process. -
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Plant Sunoort and Radioloalcal Controls 1
Although challenged by events during the RCS decontamination, the licensee applied good
radiological controls for normal activities, in response to events, and during subsequent
maintenance activities. Several industrial safety issues were noted, along with
deficiencies in the waste management of RCS decontamination water. The waste water
was controlled within the plant without releases to the environment. The licensee was -
slow to recognize and respond to heat stress conditions in some plant areas. !
The licensee measured and calculated the decontamination factors (DFs) for the RCS
decontamination and reported successful results in lowering dose rates for future work.
Radiological controls for RCS decontamination work were well planned and radiation
protection personnel maintained close oversight of work. Radiation protection was
effective in keeping workers' radiological exposures ALARA. Radiation Protection response
to the spill from LD-V-226 was adequate.
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TABLE OF CONTENTS
EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . il
TA B LE O F CO NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
REPORT DETAILS
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Summary of Facility Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1. Decommissionina Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01 Conduct of RCS Decontamination Operations . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 RCS Decontamination Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.2 Decontamination Events (VIO 98-04-01 throuah VIO 98-04-03) ....... 3 .
II. Decontamination Sunoort Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
M1
Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
M1.1 Maintenance Suocort (URI 98-04-04. VIO 98-04-05) . . . . . . . . . . . . . . 12
M1.2 Pressure Boundarv Testina . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
M1.3 Repairs of the Decontamination Pressure Boundarv . . . . . . . . . . . . . . . 16
El Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
E1.1
Enaineerina Evaluations for RCS Decontamination . . . . . . . . . . . . . . . . 17
E1.2 pecontamination Procedure Safetv Evaluation . . . . . . . . . . . . . . . . . . . 17
Ill. Plant Suncort and Radioloaical Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
R1 Radiological Surveys and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
R2 Response to Events and Radiological Challenges . . . . . . . . . . . . . . . . . . . . . . 23
R2.1 Radioloaical Saf etv . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
R2.2 Industrial Safety issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
R2.3 Waste Water Manaaement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
IV. Manaaement Meetinas ..........................................26
X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
X3 . Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
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ITEMS OPEN, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
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LIST OF ACRONYMS US ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
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REPORT DETAILS
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Summarv of Facility Activities
Special NRC inspection of the chemical decontamination of the reactor coolant system
(RCS). Specific areas reviewed included activities associated with ALARA planning,
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radiological controls, procedural controls, and process performance. The NRC reviewed
the events and problems encountered during the decontamination, and performed a special
inspection of the organizational and management issues contributing to licensee
performance:
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On August 3,1998, A.R. Blough, the Director of the Division of Nuclear Materials
Safety, toured the site and together with other NRC Region I and NRR staff met
with licensee management on the status of decontamination activities. Attachment ,
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I provides the licensee's presentation handout for the meeting.
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On August 20,1998, H. Miller, the NRC Region i Administrator, toured the site and I
met with licensee management to discuss NRC concerns highlighted by recent
events. !
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On September 3-4,1998, W. Axelson, T. Johnson, G. Pangburn, T. Frearichs and
W. Raymond conducted interviews to obtain information on the organizational and
management issues associated with the RCS decontamination.
Section X3 provides additional details of these efforts.
l. Decommissionina Operations
01 Conduct of RCS Decontamination Operations
01.1 RCS Decontamination Overview
a. Insoection Scone (71707,71801. 42700,40801. 62801,93702. 92702)
The inspection scope was to provide overview of the chemical decontamination
process using a combination of resident and region-based personnel. The
inspections were conducted during routine day-shift operations, and during weekend
and back-shift periods. Attachment ll of this report provides a list of references and
materials used during the inspection.
b. Observations and Findinos
Chemical Decontamination Overview
The licensee continued with preparations for decontaminating the primary system in
order to reduce radiation exposure during decommissioning activities. NRC review
of the decontamination preparation was provided in inspection 98-03. This licensee
used the Siemen's process called HP/ CORD D UV. The letters represent the
following: HP = permanganic acid, CORD = chemical oxida' tion reduction
decontamination, D = decommissioning, and UV = ultraviolet light. The process
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used the residual heat removal (RHR) pumps to provide about 2000 gallons per
minute (gpm) flow to circulate the. decontamination solution throughout the RCS
and portions of the RCS attached piping (the high pressure safety injection system,
the chemical and volume control system, the purification system, and the RCS fill
and drain system). The licensee installed temporary pumps [ letdown booster pump
(LDBP) and letdown return pump (LDRP)] to allow the use of the in plant
-domineralizers with an optimum flow of 200 gpm. The decontamination flow path
is shown in Figure 1.
After achieving system operating conditions for temperature and flow, the first step
in the CORD process was to add permanganic acid to soften the chromium layer.
The decontamination step begins with the addition of oxalic acid to remove the
radioactive corrosion layer. The licensee used the plant domineralizers following the
oxalic acid addition to remove the radioactivity from solution and transfer it to the
spent resin storage tank. The A and B letdown domineralizers were used, along
with the C domineralizer, which was isolated from the spent fuel pool (SFP) for the
duration of the process. Cleanup from each phase was accomplished by the use of
ultra violet light and the addition of hydrogen peroxide to remove the oxalic acid,
and using the D domineralizer to clean the system. The licensee planned that the
fourth and final cycle would be a more aggressive CORD-D process.
The licensee implemented the decontamination process with an initial plan to
complete four CORD cycles. The major milestones for the RCS decontamination
were as shown in Table 1. The inspector reviewed the licensee's preparations for
the RCS decontamination by independently observing the status of plant systems
and equipment, and verifying that the plant configuration and process controls were
as specified per procedures, and Sections 4.1,4.2,4.3 and 4.4 of SPL 10.11-1,in
particular. The inspector verified, on a sampling basis, that workers followed the
procedures during the decontamination process.
Although decontamination cycles 1 and 2 were successful relative to achieving the
desired dose reductions on piping that would be removed during decommissioning,
the plant experienced several problems and events that challenged plant personnel.
Section 01.2 and E1.2 below summarizes NRC review of the combination of
process, personnel and equipment performance problems encountered. Based on
the results achieved after cycle 2, and after considering the benefits and risks of f
conducting a third cycle, the licensee determined on August 13 that a third cycle
would not be performed. The RCS decontamination ended after CORD cycle 2.
After removing the chemicals from cycle 2, the A RHR pump remained in service
through the end of the inspection period to cool the RCS and reduce chromium
levels in the RCS water.
Decontamination Factors and Dose Reduction Factors
The RCS decontamination was successful in reducing dose rates in the RCS and
. connected piping, and in the plant areas containing reactor related piping. The
licensee planned to achieve a minimum decontamination factor of 15 from the
CORD process. The licensee achieved this goal with an average DF of 15.65. The
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licensee reported that a total of 130 curies of radioactivity was removed from the
system, which will result in an expected dose savings of about 900 person rem
over the remainder of the decommissioning.
c. Conclusions
The RCS chemical decontamination accomplished substantial dose reduction in
reactor piping; however, the process was hampered by several events that
challenged plant personnel and the safe control of the radiological source term.
01.2 Decontamination Events (VIO 98-04-01 throuah VIO 98-04-03)
a. Inspection Scooe (71707,71801,42700,4080162801,93702,927021
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The purpose of this inspection was to review licensee actions following operational
events during the RCS decontamination. The inspection included observations of
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plant equipment, logs and records; interviews of plant personnel; a review of tests, j
troubleshooting, repairs, and procedure changes; and, a review of the licensee i
assessments and event investigations. The inspector completed an independent
evaluation of each event and identified contributing causes. When licensee event
summaries and investigation results were used, they were validated by the
inspector to develop NRC conclusions and findings.
b. Observations and Findinas
Several events occurred which interrupted the RCS decontamination process. The
events included a leak of about 1200 gallons from the letdown system on July 27,
1998; a failure of the C ion exchanger and post filter, resulting in the transfer of
resin into the process piping on August 11; and, the failure of the discharge piping
in the LDBP and a leak of 125 gallons on August 13. Following each event, the
licensee suspended the decontamination cycle, took immediate actions to mitigate
the consequences, investigate the event causes with the assistance of independent
review teams as needed, addressed effected equipment and material deficiencies, l
Imptoved the decontamination process controls, and took actions to prevent I
recurrence. The response to the July 27 event was the subject of a special
management meeting at the site on August 3,1998, and a telephone briefing on
August 7 with NRR and Region i management. A copy of the licensee presentation
slides for the August 3 meeting is provided in Attachment 1. Licensee investigations
of the events and their causes were generally thorough and corrective actions were
appropriate,
b.1 July 27 Leak - Event Summary
The injection of permanganic acid started at 10:00 p.m. on July 26 to begin CORD
Cycle 1. After the chromium concentration reached equilibrium in the RCS, the
licensee injected oxalic acid at 10:30 a.m. on July 27 to begin the next step in the
decontamination process. While placing the B demineralizer in service per NOP
2.7-1 starting at 11:33 a.m., the plant operators noted unusual vibrations and noise
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from plant equipment at locations in the ion exchange alley and in the primary
auxiliary building (PAB)e The operators made two attempts to place the B
l domineralizer into service, but stopped when anomalies in header flow were noted. i
The operators conferred with the decontamination project team regarding the
possible causes and corrective measures. The operators verified the B domineralizer
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valvo lineup, enhanced venting carbon dioxide from the RCS, and throttled the LDBP
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by using the discharge valve to reduce the flow rate.
During the third attempt to place the B domineralizer in service at 12:38 p.m., a
pressure transient caused relief valve LD-RV-252 (200 psig setpoint) to lift and
recede, which caused severe piping vibrations that resalted in leaks from the RCS at
two locations in the letdown system (ACR 98-620). Plant operators followed AOP
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3.2-70 and SPL 10.11-21 to respond to the leak and shutdown the operating '
pumps in the letdown line, and isolated the letdown line by closing valves LD-MOV-
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200 and LD-V-405. Based on initial reviews of the amount of water collected by i
the aerated drain tank, the licensee estimated that about 1000 gallons of 1
- decontamination water had leaked from the RCS. The estimate was revised to '
about .1200 gallons (Reference 21 - a range of 1200 to 1500 gallons) after
considering uncertainties in the measurement.
After the letdown piping failed. the operator response to detect and isolate the leak
was good. The leakage occurred at the time when the RCS water contained the
highest concentration of corrosion products (about 0.69 microCi/ml). The leakage
caused several areas within the plant to become contaminated, and the areas were
required to be centrolled as airborne radiation areas. However, allleakage was
contained within plant buildings and was collected by plant sumps and tanks. The
majority of the leakage was directed by the pipe trench to the sump in the spent
resin tank area, which pumped the fluid to the aerated drains tank (Attachment II,
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References 31 - 35). There was no release of radioactive water to the environment. '
Radiological controls to clean up the leak and minimize airborne activity was good. !
Section R2.1 describes further NRC review of the response to tne radiological ;
conditions caused by the event.
Licensee Investiaations and Follow uo Actions
The July 27 leak occurred when the decontamination process piping failed at two
locations. The sensing tubing for letdown pressure instrument Pl 113 failed in the
purification pump area (PAB 21' 6" elevation); and, a one-half inch diameter pipe
failed on letdown line (3"-CH-151R-227)just upstream of valve LD-V-226 inside the
pipe trench. Most of the July 27 leak was caused by the failure of the LD-V-226
drain line. Several other components were affected or damaged by piping vibrations
during the July 27 event, or were discovered to be degraded during the follow up
reviews and investigations.
Pressure instrument PI-113 was replaced along with the tubing up to the isolation
l . valve. Pressure transmitter PT-113 was permanently damaged by the July 27 event
! and was not repaired. The licensee used a camera with local Pl 113 to provide for
i operator monitoring of letdown pressure from the control room. inspections and
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flow tests were completed to investigate possible sources of restriction in the
domineralizer header, including an evaluation of LD-V 238, LD-CV-343A and LD-
TCV-113A. Both relief valves in the letdown line were tested and found to lift at
the correct setpoints, nominal 200 psig and 500 psig. LD-RV-252 was replaced.
The letdown line was flow tested at 30 and 150 gpm to confirm no flow blockage
existed. The RCS decontamination flow path was modeled using a thermal-
hydraulics analysis; the results were used to improve domineralizer operations, and I
to study the system transient response. l
The LD-V-226 drain line failed when a pre-existing, through-wall flaw was driven to
failure as a result of the multiple system pressure transients and the piping
vibrations, while placing the domineralizer in service (References 25,26,27). The
licensee could not determine whether the pre-existing flaw produced visible leakage
prior to the event. The failure to include this piping in the decontamination leak !
checks was a missed opportunity to have identified degraded conditions. The
licensee removed the remnants of the failed half-inch line at the junction with the
three-inch line, and applied an engineered clamp to restore leak tightness of the
header (Reference 16). The licensee completed walkdown of lines affected by
event to assure no other structural deficiencies existed.
Several minor pin holes leaks in the purification line 2"-AC-151R-101 near valve
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WD-V-10F vere repaired by cutting out a section of affected piping and welding in
a replacement pipe. This repair also addressed other pre-existing defects in the
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purification piping, evidenced by an old, non-engineered patch covering a degraded
pipe section. The licensee established enhanced monitoring of purification piping
inside the pipe trench using additional cameras. During the remainder of the RCS
decontamination, the licensee continued to address minor leakage in the purification
system piping, including: the replacement of an elbow with pin hole leaks, the
identification of extensive pitting and flaking of the resin slurry line (ACR 98-718);
and, the identification of additional pin hole leaks on 2"-PL-152-19 near valve WD-
V-10A (ACR 98-725).
Licensee actions to address the material deficiencies, and to provide continued
domineralization for the RCS and the SFP continued at the end of the inspection
period. Refer to Section M1.2 for further NRC review of purification and letdown
piping repairs.
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Event Causes
The July 27 transient occurred as a result of a severe flow restriction (blockage) in
the letdown line while placing the B domineralizer in service. Following extensive
testing and engineering evaluations, the licensee verified that the letdown flow path
had no fixed blockage. Several evaluations (including radiographic) were performed
of the letdown postfilter inlet isolation valve, LD-V-238. The 3-inch gate valve
requires 19.5 turns of the handwheel to fully open. Immediately after the July 27
. leak, the valve was opened an additional 12 turns, which indicated it had been only
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partially open when the B domineralizer was aligned per NOP 2.7-1. This operating
procedure required that valve LD-V-238 be fully open. Based on the radiography
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and additional flow analyses, the licensee concluded that LD-V-238 had been
almost completely closed on July 27. The mispositioned valve caused the flow
- restriction which, combined with the LDBP 450 psig discharge pressure, resulted in
the pressure transient that lifted LD-RV-252 and the piping vibrations, line failures
and leaks. Subsequent problems operating valve LD-V-238 were described in ACRs98-709 and 98 723.
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The licensee concluded a primary cause of the transient was the improper response
by the operators and decontamination staff to the letdown flow anomalies, as was
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evident in the repeated attempts to operate the domineralizer without resolving the
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observed anomalies. The staff failed to exhibit a questioning attitude to resolve the
anomalies. A second major cause was inadequate engineering evaluations and
- procedure guidance for the operating configuration created for the RCS -
decontamination (i.e., the failure to adequately address the potential for pressure
transients and the over reliance on operator actions to control pressure). Following
the event, the licensee added a pressure switch to the letdown line and modified
the control circuit for the LDBP (Reference 19) to provide a high pressure trip that
assured the domineralizer piping and valves operated below 150 psig.
Contributing factors included: the inadequate communication of plant conditions
from the field to the control room; and, the focus by the decontamination team on
the need to place a domineralizer into service once the chemical had been injected,
which was not as time-critical as perceived by the plant staff. Despite consultation
with the Siemens staff, licensee actions were ineffective to integrate information
when making real time decisions during the evolution. The above issues were
examples of weaknesses in licensee control of the RCS decontamination process.
Licensee actions to assure readiness to proceed with Cycle 2 of the RCS
decontamination were summarized in DPM 98-155 (Paference 37). The licensee
revised procedures to ensure added cautions were taken when placing
domineralizers in service, and to add guidance on how to operate the LDBP. To
improve guidance for operating manual valves with handwheel extensions like LD-V-
, 238, the licensee developed a list that identified the number of turns needed to
operate each manual valve used during the RCS decontamination. The licensee
took measures to enhance monitoring of CO2 generation; enhanced communications
, and clarified expectations for the control of decontamination activities; and,
performed flow testing and additional thermal hydraulic analyses to clarify decision
criteria for decontamination operations. The inspector completed an independent
review on August 9,1998 to verify the licensee had completed the corrective
actions necessary to proceed with Cycle 2 of the RCS decontamination.
Findmas
,, The licensee completed leak checks of the RCS decontamination boundary in
i accordance with Attachment 14 of SPL 10.11-1. While the checks focused on the
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temporary connections and accessible portions of the boundary, the licensee did not
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conduct walkdowns or leak checks of portions of the RCS boundary inside the pipe
- trenches. The failure to review this piping was a missed opportunity to identify
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possible leakage in the purification piping, and the degraded drain line connected to
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LD-V-226. Plant procedures were weak in this regard (see Section M1.2), and this
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is the first of two examples of a violation of Technical Specification 6.8.1 relative to I
SPL 10.11-1 (VIO 98-04-01). In response to NRC reviews during the August 3, i
1998 manageme'nt meeting, the licensee conducted a walkdown of the complete i
i RCS decontamination boundary.
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Section E1.2 of this report describes the licensee safsty and technical reviews for
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the RCS decontamination. NRC review identified several engineering support
weaknesses, which contributed to deficiencies in the decontamination process
controls, as defined in the Master Decontamination Procedure, SPL 10.11-1, and
associated procedures for operation of the letdown system. Specific inadequacies
in the evaluations for the modifications that installed the LDBP included: the failure
to adequately address the operational considerations related to the differences in
pressures between the LDBP and the domineralizer components, particularly in
regard to transient conditions; and, the deficiencies in the avaluations related to the
pressure testing of components in the decontamination boundary. As a
i consequence, the decontamination procedures did not adequately address potential
transient conditions in the letdown system. Although SPL 10.11-1, Sections 6.4 I
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and 6.8, and SPL 10.11-24, Section 6.3, provided instructions for operation and l
static system pressure control using the letdown booster and return pumps, there
was no guidance for transient pressure response, or termination criteria for
abnormal pressure and flow conditions during domineralizer operation. As noted
above, the licensee also concluded that the failure to address the potential for
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transients contributed to the July 27 leak. This inadequacy in SPL 10.11-1 and
associated procedures was a second example of a violation of Technical
Specifications 6.8.1 (VIO 98-04-02).
NRC concerns regarding valve lineups and plant configuration control were
addressed in telephone conferences on July 8 and 15,1998. Licensee corrective
actions to address deficiencies in this area were described in a letter to the NRC
dated July 16,1998, and included basic valve operation training for personnel who
- mr.nipulate valves or verify valve positions. Inspection 98-03 addressed the
- scensee's failure to follow procedures in the mispositioning of LD-V-238. The
corrective actions implemented prior to the start of the RCS decontamination were
not effective to preclude the events that contributed to the July 27 leak. The
continuation of configuration controlissues and the faliure to take adequate
corrective actions to address weaknesses in configuration control were examples of
inadequate corrective actions.
b.2 Auoust 11 Domineralizer and Filter Failure - Event Summarv
Following repairs to the domineralizer system, the licensee completed the cleanup
from CORD Cycle 1 using the C domineralizer (C IX). New resins were placed in the
C IX on August 10 in preparation for CORD Cycle 2 of the RCS decontamination.
The licensee initiated the injection of permanganic acid to begin Cycle 2 at 1:35
a.m., and injected oxalic acid at 10:01 a.m. on August 10,1998. At 3:55 a.m. on
f
d
7
.- - , - _ . _ - - - ,, . - _ - _ _ ._- .
-. . - - . - . - . . . . - - .- - - - - - - .-- . - - -
l
August 11, radiation monitor RMS-35 came into alarm, indicating an increase in ;
e
PAB hallway area dose rates from 0.5 mrem /hr to 5 mrem /hr. .
1- Licensee reviews to identify the cause of the increased radiation continued on
i
. August 11. Initially, the licensee considered and then discounted whether the !
'
addition of peroxide had caused the increase. After identifying that domineralizer
' flow data had been incorrectly recorded, the licensee concluded that radiation levels
' increased following the actions to increase flow through the C domineralizer from
65 to 90 gpm per SPL 10.1-48.~ About 30 (of 45) cubic feet of resin passed
through the lower retention element of the C IX at the elevated flow rate. The i
radiation levels were caused by the movement of activated resin from the C IX,
which had migrated into the piping that was part of the RCS decontamination
boundary. The C IX postfilter, FL-66-1 A, failed to retain the resin.
Radiation surveys confirmed elevated dose rates existed in portions of the high
pressure safety injection (HPSI) piping from the PAB to the SI-861 valves (A
i
through D) inside the containment. Contact dose rates on the lower end of the C IX
were measured at 40 R/hr prior to the event, and had decreased. Dose rates were
as high as 5 R/hr on contact with the piping inside the containment, with general
area dose rates at 1 to 2 R/hr at 30 cm from the line. The event might have been
worse had the filter failed prior to the last C IX change out when dose rates were
. about 250 R/hr. The affected portions of the HPSI piping inside the PAB were in
the pipe trench, and were not readily accessible to workers. The affected HPSI
piping inside the containment was already controlled as a high radiation area
.
4
because of the RCS decontamination process. The licensee isolated portions of the
-decontamination flow path to keep the resin from entering the RCS piping.
Licensee investiaations and Follow uo Actions
.
The failure of the C IX eliminated its use for the remainder of the RCS
decontamination. The licensee used the letdown domineralizers to complete the
!
RCS decontamination, and developed alternate methods for use of the RHR system
in the purification mode (Reference 45). The licensee opened the top of the C IX
and used boroscope examinations to investigate the internals. This examination
identified a failed retention element. Since the C IX was the only source for SFP
clean-up, licensee efforts continued at the end of the inspection period to provide a
means to purify the SFP. The licensee was considering actions to accelerate the
implementation of phase 2 of the nuclear island modifications, which would install
new SFP domineralizers in the spent fuel building. The licensee flushed the
decontamination chemicals out of the HPSI piping per procedure SPL 10.11-10
(Reference 9). The activated resins remained in the HPSI system pending the
development of a procedure using a vendor-supplied process to flush the resins out
of the piping.
Postfilter FL-65-1 A was opened and inspected (Reference 39). The screen material
included in the retention element had disintegrated in the presence of the
4
decontamination fluid, allowing the resin to pass through the filter. The element
that failed had been inst 611ed by the licensee on July 25,1998 as a temporary
.
8
4
- - - -
v--
. .-, ..- -- .....- - . - . - - . ~ - . ~ . - ~ - . . . - - - . .
.
- modification (References 41,42). The normally insta! led, vendor-supplied 1 micron
<
filter element was replaced with a 150 micron element fabricated onsite using the
materials from the Haddam Neck and Millstone warehouses. Post-event chemical
tests on the screen material used in the modified element showed that it had a high
concentration of iron, even though the stock code indicated the material was type
304 stainless steel. Further ucensee investigation determined that the material was
stainless steel, but the screens had dissolved because of the fine mesh used in the
fabric of the filter element. Licensee evaluations continued at the end of the
inspection period. Even though the technical evaluation (Reference 42) addressed
the structuralintegrity and hydraulic performance of the modified element, it was
inadequate in that it did not consider whether the new element would be compatible
with the decontamination fluid. This was an example of a weakness in engineering
support for the RCS decontamination.
. Findinas
,
The technical evaluation in support of Jumper 98-36 (Reference 42) noted that the
lower retention element passed resin with system flow rates at 140 gpm. The
licensee concluded that flow testing had established proper C IX performance with
flows at 100 gpm, which was established as a limit for domineralizer operation.
Procedure SPL 10.1-48, Step 1.6.2, initially provided guidance to adjust purification
system flow through the C IX at the discretion of the Decontamination Manager,
but was revised on July 25 to limit flow to 100 gpm (TPC 98-117). Ultimately, the
combination of technical evaluations for Jumper 98-36 and procedure controls was
inadequate to assure the activated resins remained in the C IX during domineralizer
j
operation. The failure to provide adequate procedures for purification system
operation during the RCS decontamination was an example of a condition contrary
to Technical Specifications 6.8.1 (VIO 98-04-03).
4
The licensee had identified that the lower retention element in the C IX was
potentially failed and not functioning properly, since the postfilter had been
inspected and cleaned of resins three times from July 14 to July 25,1998
(Reference 40). The degraded condition of the retention elements was noted in the
technical evaluation for Jumper 98-36, but the option of replacing the retention
elements was rejected due to "ALARA considerations." Although the degraded
condition of the retention element was noted, the failure to adequately address this
' condition centributed to the migration of activated resins and the loss of control of a
radiological source term. The failure to take timely action to address a degraded
material condition was an example of inadequate corrective actions.
'
b.3 Aunust 13 Letdown Booster Pumo Leak - Event Summary
'
During the conduct of CORD cycle 2 on August 12, the LDBP tripped resulting in a
level transient in the volume control tank. The operators responded per AOP 3.2-70
to control the transient and isolate the letdown line. The licensee identified and
i_
replaced a blown fuse in the LDBP control circuit of the (ACRs98-679 and 98-684).
The fuse blew because vibration of the skid mounted control panel caused wires to
)
fray resulting in an electrical short circuit.
! 9
-- . -. -. -. - ,
. -- - . _ . - _ _ _ ,
l l
l
The LDBP was started at 4:05 p.m. on August 13 to resume cleanup for Cycle 2.
Operators noted excessive vibrations in the LDBP and requested engineering sunrort
to measure the vibrations levels and evaluate the pump status. A technicai support
engineer conducted measurements between 5:45 pan, and 6:00 p.m., noting
vibration levels 1.2 inches per second (ips) on the coupling,1.7 ips on the i
baseplate, 2.5 ips and 4.0 ips on the motor, and 6.0 ips on the skid mounted I
control panel. The vibration, levels were greatly increased from the baseline levels I
taken when the LDBP was initially installed (measured at much less the 1.0 ips).
The technical support engineer also noted excessive movement of the LDBP skid.
'
The technical support engineer immediately informed me duty shift manager and the
lead representative in decontamination projects group of the excessive vibrations,
and recommended that actions be taken as soon as possible to tighten the
tumbuckles to lessen skid vibrations and allow further evaluation.
i-
!
! Before the licensee took additional action, the piping in the discharge of the LDBP
l failed at about 7:09 p.m. on August 13, resulting in a spill of about 125 gallons of
decontamination fluid in the lower level of the containment. Plant operators j
l recognized the indications of a RCS system leakage based on inputs to the
l containment sump, and promptly responded to isolate the leak per AOP 3.2-70 and
- SPL 10.11-21. The health physics group responded to :ntrol the spread of
,
'
contamination and minimize the potential for airborne radioactivity by flushing the
spilled water back to the containment sump.
Licensee Investiaations and Follow uo Actions
Subsequent investigation of the LDBP identified the following damage: the socket
-
weld at the transition from the 3 inch di%harge line to the 1 inch recirculation line
had failed, with indications of a 180 degree crack; the pump to motor coupling was
i damaged; one of four motor mounting bolts on the cart was failed; and, one of four
! turnbuckles was bent. The failed weld was the source of the leakage. The licensee
l determined that the weld failed due to high cycle fatigue caused by the excessive
i vibrations. Licensee actions were appropriate to repair the damaged equipment,
!
and investigate the structural integrity of the pump skid and associated piping
(References 5,6) prior to the resumption of decontamination operations on August
1
18,1998. The piping inspections and non-destructive examinations were thorough
l to verify system integrity and that there were no other vibration induced defects.
!
l
l- The licensee also concluded that the high vibrations on the LDBP skid had
l contributed to various problems experienced with the LDBP control circuit during the
, RCS decontamination, including the start circuit problems, frayed wires and blown
i
fuse. The licensee operated the pump discharge valve almost fully throttled closed
in order to control the pressure and flow in the down stream letdown piping. This
mode of operation contributed to the vibrations and " rough" operating conditions on
the skid, as evidenced by the tendency of the LDBP discharge valve to vibrate open.
The licensee response was to install a restraining device on the discharge valve.
i
!
, 10
l
!
!
- _ _
.
_ __ -. . ._ _ . _ _ _ _ . _ . _ . _ . _ _ _ - _ _ _ _ . . . _ . . _ . _ _ . _ . _ _ _ _
Findinas
The LDBP f ailed in service, resulting in a 125 gallon leak to the containment sump.
There was no leakage to the environment. The operator response to detect and
isolate the leak was good. HP follow up to clean up the leak and minimize the
spread of airborne activity was good. ,
The LDBP was installed per MMOD CY-98509, DCY-OO-OO54-98 and DCY-01-
0054-98. The original plan was to use anchor bolts to mount the pump skid
directly to the containment floor. The plan was changed by leaving the skid
mounted on wheels to facilitate installation and removal. Turnbuckles were used to
restrain the skid at the four corners of the cart. No lock nuts were used, and the
turnbuckles loosened from the vibrations that occurred when operating the LDBP.
Engineering support was inadequate regarding the review to change the mounting
details, in setting the design and operating limits for the LDBP, and in how the LDBP
operation was integrated with the RCS decontamination. Section E1.2 below also
addresses this area.
Engineering support was inadequate on August 13 by not recommending immediate
shutdown of the LDBP on the basis of vibration data that was recognized by the
support staff as excessive and unacceptable. Although vibration measurements
taken at several locations on the pump skid were found excessively high, there was
no official procedure or acceptance criteria used by engineering to evaluate the
LDBP for continued operability. Prior to the pump failure, the operator evaluations
of the vibration did not exhibit a good questioning attitude; operator actions were
not adequate to preclude the LDBP failure.
The failure to take prompt actions to shutdown the LDBP when excessive vibrations ;
were observed was an example of inadequate corrective actions. Continued pump i
operation resulted in the piping failure and the loss of control of the radiological
source term. Licensee communications and actions to integrate information from
the decontamination and support staff were ineffective when making real-time
decisions during this event. This was a second example (see Section b.1) of a
weakness in the control of the RCS decontamination.
c. Conclusions
Licensee preparations for and conduct of the RCS decontamination was generally ,
adequate, but many equipment and personnel performance issues and process
deficiencies were evideret in several events that challenged the plant staff. The
operators and support staff responded properly to major events (a 1200 gallon leak,
and the loss of control of resins and radioactive material, and a 125 gallon leak) to
'
mitigate adverse radiological conditions, and take appropriate actions to address
degraded conditions. However, weaknesses were noted in the incomplete leak
checks for the RCS decontamination boundary, several deficiencies in engineering
evaluations and procedures, and in the failure to take effective corrective actions for
adverse conditions, such as excessive pump vibrations, failed domineralizer
11
_ , -_ -_-
__ __ _ . - _ - _ _ . _ _ _ _ . . _ _ . _ . . _ _ _ _ . - _ _ . _ _ . . . . _
!
retention elements, and inadequate control of the plant configurat;on (valve lineups).
Corrective actions were not comprehensive to preclude additional events.
NRC staff concerns regarding inadequate procedures and corrective actions, and in I
.
the control of shutdown operations, were previously noted in NRC Inspections '
96-10,96-11,96-80 and 96-201, ar.d in the May 12,1997 Escalated Enforcement i
.,
Action. In addition to root cause and event investigations for the individual events, '
licensee actions continued at the end of the inspection to address organizational
issues, and to complete a common cause evaluation for the RCS decontamination
events.
IL pecontamination Support Activities
M1 Conduct of Maintenance '
' M1.1 Maintenance Sunoort (URI 98-04-04. VIO 98-04-05)
a. Insoection Scooe
1
Using Inspection Procedure 71707,61726 and 62707, the inspector reviewed plant
maintenance and surveillance in support of the RCS decontamination.
,
b. Observations and Findinas i
i
'
Several equipment problems hampered the decontamination process and challenged
plant personnel. Some problems involved valves and filters that support the plant
purification process, which exhibited poor material conditions that were pre-existing
and became apparent as the decontamination process proceeded. Some problems ,
occurred as a result of rough equipment operation or system transient conditions. !
Plant workers responded to each problem to complete repairs. While work was
conducted with good regard for equipment and personnel safety, exceptions to
good perfortnance were noted, as discussed below,
e repair of several valves that failed on demand (diaphragm type)
e repair of pressure instrument and drain line from July 27 leak
e
'
repair of LDBP discharge line following August 13 leak
. e purification line 2"-AC-151R-101 pin hole leaks
,
o waste disposal line 2"-PL-152-19 (WD-V-10A) pin hole leaks
e VCT manway cover leak (ACR 98-668)
e C lon exchanger and RCS letdown postfilter replacement
e resin slurry line corrosion (pitted, crusted, flaking)
e LDBP control circuit problems - frayed wires, start problems, blown fuse
e testing and replacement of letdown relief valve LD-RV-205,252
Eurification (Demineralizer) System Leaks and Valve Performance
i
The licensee encountered several problems with the domineralizers and associated
piping and valves during tne RCS decontamination, which challenged personnel and
<.
, 12
,
-- --
-. ._ _-- . _ _. .- . . - . - _ - .
_
- - _ . . - - - - - - . _ -
hampered the smooth conduct of the process. Although the licensee had identified
the potential vulnerabilities on the use of diaphragm valves, the plant experienced
multiple valve failures during domineralizer operations. There was no preventive
maintenance on the demineralizer system valves; valve maintenance was conducted -
as failures occurred. The domineralizer valves are located in the pipe trench, but are
operated by long reach rods. Most valves with reach rods have no design provision
for providing visual feedback on valve position. The valves are often hard to
operate, and provide ambiguous tactile indication that the valve has reached the end
of its travel.
The plant design included the use of thin-walled stainless steel piping ("speedline
. pipe") associated with the domineralizers. The nominal wall thickness for the
Schedufe 10 piping was 0.10 inches. The licensee identified severai leaks in the
piping during the RCS decontamination, and continued to discover leaks after the
decontamination (ACRs98-636,98-639,98 687,98-725, and 98-767). The resin
slurry header near the D domineralizer had external corrosion and was bulged in
spots (ACR 98-718). Licensee evaluations determined the piping had pin hole
defects resulting from stress corrosion cracking type indications. .Since the licensee
had not inspected the piping during the preoperational leak checks, it was
indeterminant whether the leaks had existed prior to the addition of the
,
decontamination fluids. However, the RCS decontamination fluids likely removed
wall material in the flaw areas and initiated the leaks.
There is some evidence that degraded conditions existed in the purification system
piping prior to the RCS decontarnination. A temporary patch (non-engineered damp
consisting of rubber wrapping and two common hose clamps) was identified on July
30,1998 during reviews inside the pipe trench following the initial identification of
leaks. Maintenance personnel had no record on when the patch was installed, but
estimated it had been in place for many years.
Failed Riacina - PAB Floor Block
The licensee used the overhead crane in the PAB to lift floor block RS-5 to gain
access to the pipe trench to repair valve LD-RV-252 (Reference 18). The block
weighs about five tons, and was handled per procedure WCM 2.2-7, PAP / Pipe
Trench Floor Block Lifting Procedure. While WCM 2.2-7 provides general guidance
to safely handle the load, the procedure provided no details on how to rig the block.
Plant workers, after consultation with plant engineering and the safety personnel,
rigged the block using two turnbuckles and two metal slings. When the block was
lifted and moved east on August 2,1998, the northwest attachment failed, causing
the corner of the block to fall two inches and strike block RS-4 (ACR 98-646). The
attachment failed when a one inch diameter swivel eye bolt pulled out of the
Richmond Rocket insert in the concrete. No personnel injuries occurred and
licensee engineering verified that the structural integrity of block RS-4 was not
compromised.
The licensee conducted an apparent cause for the event, in recognition that the
failure to control the heavy load was a significant precursor event with the potential
13
,
, , . . g - . . _ . - ., , _ - , . -
,,
- . - . - - - - . - - - - - . - _ - - - . - . -.
I
I
to adversely affect systems and components below the 21-foot elevation of the
PAB. Although engineering and maintenance personnel verified that the lift angle
was acceptable (31.4 degrees versus a minimum acceptable of 30 degrees), neither
the workers nor the engineering or safety personnel verified that the turnbuckles )
were used properly with the slings; i.e., configuring the turnbuckles on the same I
side of the load, rather than being diametrically opposed. Rather than spreading the 1
load evenly over four eye bolts as intended, the swivel insert became over stressed
and failed when the weight of the block was spread over two bolts.
There were several contributing causes to this event. The workers made a new i
rigging method for the RS-5 lift because the old lift rig, which had been used
successfully for years, had been removed from service in 1996 due to no formal
.
tests to assura operability (ACR 96-399). The lift rig remained stored in the east
and of the PAB within 50 feet of RS-5. Alternate rigging was available at Millstone,
.
but was not obtained; instead, the plant workers created a new rigging method
using available materials at the CY site while trying to proceed with the repairs.
Although the workers and support personnel resolved questions that were raised
regarding the lift angle, not all questions were resolved regarding the use of
turnbuckles. Both engineering and plant safety personnel approved the intended lift
method without seeing the rigging installed. Communications between maintenance
and engineering was poor regarding the configuration used for the turnbuckles:
engineering assumed the workers knew how to install the rigging properly; and the
workers believed the support personnel had approved the configuration. The
activity was completed under schedule pressures, which was a factor on how well
the licensee prepared for the lift and reviewed the lift method.
Licensee actions continued at the conclusion of the inspection to address the
' apparent cause recommendations, and to recover block RS-5. The apparent cause
investigation for the event (Reference 38) was thorough and identified several
actions to improve rigging practices at Haddam Neck, including the evaluation of
training and obtaining a proper lift rig for PAB floor blocks. This item is unresolved
pending the completion of licensee actions to address weaknesses in this area (URI
98-04-04).
Although WCM 2.2-7 provided general guidance to safely handle PAB floor blocks,
the procedure provided no details on how to rig the block. The combination of
l
procedure guidance and worker training was inadequate to properly rig the block.
l
This failure to provide adequate procedures is an example of a violation of Technical
Specifications 6.8.1 (VIO 98-04-05).
i
c. Conclusions
The material conditions of purification and domineralizer system valves and piping
were poor, which hampered the decontamination efforts and challenged plant
personnel. Past NRC inspections noted previous weaknesses in this area (reference ,
NRC Inspections 96-10,98-11, and 97-01). The licensee did not adequately assess
or compensate for the material conditions in the purification and domineralizer
system until failures occurred that challenged workers and the safe control of the
.
14
.
.- -, -,- - -
. , _
- 1
. _ _ . . _ _ _ . _ _ _ . _ _ _ _ _ _. _ __ __ _ _ . _ _ _ _ _ _ . _
l
radiological source term. This was a weakness in the preparation for the RCS
decontamination. Weaknesses were also noted in the conduct of rigging for a
heavy load (PAB floor block); communications and the integration of support by
engineering and the plant safety group were poor.
1
M1.2 Pressure Boundary Testina
'
a. Insoection Scone
The inspector reviewed pressure test requirements specified by CYAPCo to assure
leak integrity of the chemical decontamination pressure boundary. Test
requirements for selected minor modifications, system repairs, and the " Master
Decontamination Procedure," SPL 10.11-1, were reviewed.
b. Observations and Findinas
Minor Modifications
The inspector reviewed selected minor modification packages listed in Attachment 11
of this report and interviewed engineering personnel. The inspector found the
pressure test required met the requirements of the RWQA program and ANSI B31.1
and provided ample assurance of the pressure boundary integrity. Adequate
technical justification was noted for the selected hydrostatic test pressures, the
majority of which were 150% of system design pressure. If a hydrostatic test was
not performed, pressure testing to a lower value had been adequately justified and J
complied with ANSI B31.1.
System Reoairs
The required pressure tests for repairs to the instrument piping, temporary patch on
- the letdown line, and the piping replacement in the SFP purification line were noted
to provide adequate assurance of the repair's leak integrity and met the
requirements of ANSI B31.1.
!
The additional piping inspections and pressure tests, performed to identify additional
locations of system degradation and/or leakage, were found to be adequate in the
areas performed. When the extent of the effort was questioned by the inspector,
CYAPCo identified their intent to expand the effort to encompass the entire
decontamination pressure boundary.
Master Decontamination Procedure Leak Checks
The Master Decontamination Procedure, SPL 10.11-1, required visual checks per
attachment 14 be performed to identify external or intersystem leakage prior to
heating up, at 2OO'F, and at a minimum of four-hour intervals once the chemical
. decontamination had begun. The visual inspections were performed by walkdowns
and remote CCTV monitoring systems. Based on review of attachment 14 and
15
. .. ~~ -._ - . . - . _ ~ . . - . . - - - . .- - . - . - - _ . . - - . . .
l
4
) discussion with plant personnel, inspection of the permanently installer' piping was
not a significant attribute of the visualinspection.
.
The lack of a thorough inspection of the permanently installed piping was a missed
'
opportunity to identify potentially pre-existing piping degradation and leakage.
Although it is not known if active leakage from the SFP purification line, relief valve i
.
LD-RV-252, or the drain line containing valve LD-V-226 (which later failed) existed {
prior to the spill event on July 27,1998, the presence of an old temporary piping !
repair, made using sheet rubber and hose clamps, could have been identified.
c. Conclusions
{ The required pressure tests provided adequate assurance for the leak integrity of the
systems and temporary equipment installed for the chemical decontamination of the
reactor coolant and associated systems and met the requirements of the radioactive
i
waste quality assurance program and ANSI B31.1. Testing specified for the repairs
to the permanent piping was found to be adequate, meeting the requirements of
ANSI B31.1. Though not technically required, the inspector noted an opportunity to
identify potentially pre-existing system degradation and leakage that was missed ,
i
due to the lack of a thorough visual inspection of the permanently installed piping !
during the operational leak checks of the Master Decontamination Procedure.
,
M1.3 Reoalts of the Decontamination Pressure Boundarv
,
a. Insoection Scoos I
i
The inspector reviewed the intended repairs for the piping failures identified during
cycle 1 of the decontamination process.
b. .Qbservations and Findinas
,
i
CYAPCo stated in part, permanent repairs would be made to the pressure boundary
of syste.ms defined as " operable" or "available" in accordance with the applicable
code requirements. Temporary repairs, determined to be acceptable based on an
3.
engineering evaluation, would be made for systems categorized as " lay-up" but
.
returned to service for the decontamination process prior to being declared
" abandoned." An acceptable temporary repair may be considered for systems
defined as " operable" or "available" based on the safety significance and expected
remaining service life of the system.
<
<
The repairs to correct the system leaks and piping failures from cycle 1 were found
technically adequate. The repair specified for the degraded piping in the SFP
purification line (categorized as " operable") required the piping to be permanently
replaced to meet code standards. The temporary repair applied to the letdown line
(categorized as " lay-up") where the LD-V-226 drain line broke off, was
.. accomplished using a pipe clamp with a pressure and temperature rating exceeding
the system requirements and a gasket material compatible with the environmental
1
conditions produced by the decontamination process.
J
4 16
- _-- , - ,. _, . , , _ .
. . - - . ~ - -- - -- - - - -
Y
,
1
c. Conclusions 4
The repairs specified to correct the pressure boundary piping failures noted during
cycle 1;were technically adequate. The graded approach employed by CYAPCo, in -
regard to the extent of repair to be performed, adequately balanced the safety
-
significance of the system and remaining expected service life prior to abandonment
with the concern for unnecessary personnel exposure, while also ensuring that the
repair was technically acceptable.
E1 Conduct of Engineering'~
E1.1 Enaineerina Evaluations for RCS Decontamination
- a. Inspection Scone (37801. 42700. 37700. 40801)
The inspector reviewed the conduct of engineering activities this period that '
supported decommissioning planning and implementation.
b. Observations and Findinas
Proto-Power Thermal-Hydraulic Model
The inspector reviewed selected portions of the thermal-hydraulic model developed
by Proto-Power for the decontamination evolution for indications of potential system
over pressure conditions. Review of data from the thermal-hydraulic model
identified the expected pressures for the areas in which the piping failures occurred
were all within the allowable pressure rating for the system. Additional modeling
performed subsequent to the spill on July 27,1998, verified the' installed system
relief valves (200 psig and 500 psig) each were capable of relieving the flow
capacity developed by the LDBP. The model also identified the maximum obtainable
pressure from the LDBP was 511 psia.
c. Conclusions
Adequate engineering evaluations and thermal-hydraulic modeling were performed
.
for the static and expected operating conditions to ensure the systems were
capable of supporting the anticipated operating parameters (e.g. pressure,
temperature, and flow) during the decontamination process.
.E1.2 Decontamination Procedure Safetv Eve!ustion
a. Insoection Scone (37801)
An inspection was performed on Safety Evaluation No. SY-EV 98-OO21," Master
' Decontamination Procedure", prepared by the licensee to evaluate the activities and
modifications' associated with the RCS chemical decontamination. The evaluation
was dated June 30,1998. Two of the supporting plant modifications, MMOD CY-
98509 and MMOD CY-98512, were also inspected.
17
_
. . . .-. - _ _ _ - . .-_ _ ___. __ _ _ ._ _ _ _ . . _ _ _ _ _ _ _
l
J
!
b. . Observations and Findinos
<
The purpose of the decontamination was to reduce the occupational doses to
personnel performing decommissioning activities by about 1000 rem. Major
,
portions of the RCS, chemical and volume control system (CVCS), and RHR
systems were included in the decontamination flow circuit. The reactor vessel and l
F internals were bypassed and not included in the flow circuit. -
4
-
Plant modifications installed to accomplish the work were identified in the safety
evaluation. The modifications included connections to the decontamination
vendor's equipment, two additional pumps, a jumper from the SFP domineralizer to )
the high pressure safety injection (HPSI) header, removal of the letdown orifices,
electric power connections to temporary equipment, and a nozzle dam and jumper
arrangement to bypass flow around the reactor vessel internal surfaces.
Two of the supporting modification design packages were inspected. MMOD CY--
985091nstalled a booster pump to provide flow to the letdown system. The LDBP
was located to provide 200 gpm to the A, B, or D ion exchangers in the CVCS
system. Shutoff head of the pump was 465 psig. Operational considerations were
not provided in the booster pump design package, instead, those considerations
were stated as being addressed in the Master Decontamination Procedure (MDP)
safety evaluation. The MDP safety evaluation stated that the booster pump was
designed to provide 200 gpm at 400 psig discharge pressure. No safety evaluation
was performed for the booster pump modification. The applicability review done to
determine whether a safety evaluation was necessary stated that, because the
systems involved had been classified as "layup" or " abandoned", changes to the
system are not considered a change to the facility as described in the SAR.
MMOD CY-98512 removed the letdown orifices and associate air operated block !
valves and replaced them with a pipe jumper. Removal of the orifices was I
necessary to allow sufficient flow through the letdown system to accommodate the
decontamination process. The design package stated that the RCS and supporting
systems would operate at less than 100 psig and 200 *F, and that the expected
maximum operating pressure and temperature for the decontamination were 165 I
psig and 200 'F. The design flow rate'of domineralizers A, B, and D was stated as '
150 gpm. In a fashion similar to the booster pump modification, operational
considerations were not addressed, and no safety evaluation was performed. The
MDP safety evolution was referenced for operational considerations. Invocation of
the system categorization as "layup" or " abandoned" justified the decision not to l
perform a safety evaluation.
The letdown orifice removal modification design package stated operating pressure
and flow parameters that were less than the expected pressure and flow delivered
from the booster pump modification. The MDP safety evaluation did not address
the apparent contradiction between the two modifications. However, the MDP
safety evaluation stated that the booster pump design pressure was well within the
piping classification assigned to the letdown line in the area. The MDP safety
evaluation noted that the letdown line had considerable pressure drop, and that
18
_
._ _ . - . _ .- -- -
- . . - . . . _ - .. . - . . ~ . -. . _ - . -- . - - . - -.
i
1.
k
there was pressure relief to the volume control tank (VCT) to maintain design
,
conditions. The MDP' safety evaluation did not address the apparent mismatch of
using a pump with a 400 psig discharge pressure to feed a system wiiii a pressure
relief valve set at 200 psig.
,
Both modifications involved welds subject to the requirements of Haddam Neck
.
specification SP-ME 925, which required dye penetrant and radiographic
examination of the welds. However, ANSI B31.1 required only visual inspection of
welds made on pipes used for service at conditions below 350*F and 1025 psig.
The provisions of ANSI B31.1 were applied to the welds. The licensee justified that
decision on the basis that the systems involved were classified as "layup", the
piping was for temporary use during the decontamination, less inspection reduced
3
occupational exposure, and that a hydrostatic test would be performed after
installation.
.
4
The MDP safety evaluation incorporated operations procedures by reference.
Several potential malfun<:tions were also considered:
A
- A break in the return line from the SFP domineralizer
- Internal system leakage
- Loss of power (1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> allowed before corrosion becomes a concern)
,
>
- Pump malfunction (loss of fluid circulation capability by a pump)
- Heater malfunction
- * Operator error
- Nozzle dam failure
- Resin spill
- Containment vent malfunction
- Leak detection and response to leaks
,
The break in the return line from the SFP domineralizer, if it occurred, was assumed
to release 1600 gallons of domineralized decontamination fluid onto the ground in
i
the vicinity of the refueling water storage tank (RWST). The basis of the
-
assumption was that operators would secure flow from the purification pump within
i 10 minutes. Yard drains in the area were plugged as a precaution against releasing
the water to the environment. Although not mentioned in the safety evaluation, the
licensee placed a dike around the area to contain a leak, if it occurred. The licensee
' conservatively assumed no reduction in activity as the solution went through the
demineralizer, and determined that an airborne release from the postulated spill was
bounded by the resin fire accident scenario,
i
I
Pressure testing and leak testing were addressed by the MDP safety evaluation.
} The mechanical modifications made to enable full system decontamination were
.
hydrostatically tested as part of the installation sequence. The decontamination
i
vendor supplied pressure test certifications for their equipment. Pressure testing of
existing plant equipment, other than components coincidentally included in the test
,
boundary of the modifications, was not addressed in the MDP safety evaluation.
'1
i 19
.. , -_ . ,- , . . ~
.. ~ - - - . - - - _ . . -. - . . - . - - . . . - . - - . . - . - - _ - -
A leak test was specified for the entire system as part of the startup sequence for
the decontamination. System flow, temperature, and pressure were held as
constant as possible. The test monitored water level in the VCT and pressurizer to
determine if leakage was present.
The wastes generated were evaluated for compliance with 10 CFR Part 61 waste
disposal criteria. In addition, the MDP safety evaluation stated that resin wastes !
!
would be sampled and analyzed prior to preparing the resin for shipment to a waste
disposal site to verify compliance with Part 61.
System material compatibility with the decontamination chemicals was reviewed by ;
the decontamination vendor. Based on the corrosion rate of SS-304,the expected
{
maximum penetration after 5 decontamination cycles was 1,24 micro meters ( m),
considered to be a negligibly small fraction of the available wall thickness of
stcinless steel piping. The corrosion rate of the inconel steam generator tubes was
higher, and up to 10 m penetration was expected. The penetration was
considered acceptable based on the 62 pm thickness of the tubes. No material
incompatibilities were identified in the safety evaluation.
No unreviewed safety questions were found by the licensee. A Technical ,
Specification change was required to eliminate RCS chemistry limits that would l
prohibit introduction of decontamination chemicals. The required change was part
of License Amendment No.193, Defueled Technical Specifications, issued on June _
30,1998. l
!
Assessment
Passing greater than design flow through the domineralizers increased the pressure
drop across the resin beds. When combined with an unexpected flow blockage, the
letdown system pressure could increase to the discharge pressure of the LDBP.
Because the lower setting of the system relief valves was 200 psig and the pump
shutoff head was 465 psig, a potential existed to lift the relief valve if a flow upset
occurred. This potential was not recognized la the safety evaluation for the
decontamination procedure, and may have contributed to the leakage event
experienced on July 27,1998.
The MDP did not address effects of potential corrosion of plant systems placed in
layup condition after the plant permanently shut down. The material condition of
the plant may have degraded in the last two years. The system leak testing
performed during the startup sequence for the decontamination did not include
walkdowns of plant systems to determine if leakage too small to detect from tank
level changes might have been present. Such leakage could be a precursor to a
large leak. The evaluation did not adequately consider the use of the
decontamination chemicals with the thin-walled (schedule 10) piping in the
purification system, and failed to assure that the purification system would remain
leak tight for the duration of the RCS decontamination.
20
. -- - . _ _ - .
.
_ . _ . _ .~ -.. _ . _ _ _ , _ _ ___ _ ._.._. _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _
,
4! The 10 CFR 50.59 applicability review performed for MMOD CY-98509, Letdown
!
Booster Pump, states that the activity does not make a change to the facility as l
I- described in the UFSAR because the system has been dep!cted in the UFSAR in the
'
"Layup" category. The applicability review further states that UFSAR Chapter 1
[
j
states that changes "Layup" systems are not considered changes to the facility
since any recategorization or modification of these components will not result in an
l unreviewed safety question. UFSAR Chapter 1 provides definitions of the various
!
' categories of plant systems, including the "Layup" category. However, the UFSAR
does not state that changes to systems categorized as "Layup" are not considcred
.
changes to the facility.
I
L c. Conclusigag
}
'
Severalinadequacies were noted in the engineering support for the RCS
decontamination, and the associated safety evaluations.
l
j lli. Plant Suonort and Red!alaa!ce! Controls
j R1 Radiological Surveys and Controls
!
j. R1.1 Radioloaical Survera
i'
a. Insoection Scoos (83750)
!
l- The inspectors reviewed the licensee's methods for determining the effectiveness of
j .he RCS decontamination through the calculation of DFs.
i
l b. Observations and Findinas
i
The licensee developed a procedure to determine the DF and monitor the
i
effectiveness of the RCS decontamination. The DF was determined by measuring
i
the exposure rates at various fixed locations before and after the decommissioning
activity. The DF was calculated by dividing the initial measurement by the final
measurement. ,
'
i
4
The licensee monitored over 40 locations (pipes and components) with
j. teledosimetry units that provided remote read-out capability. This enabled the
j licensee's staff to monitor the exposure rates without actually entering the areas
'
and avoided potential radiation exposure to workers. However, for the initial
(baseline) and final calculation of the DF, radiation protection technicians went to
i. each location and performed a radiologicaldose rate measurement with a hand held
t instrument.
.
1 The licensee expected to obtain an average DF in excess of 15 for the RCS
.
decontamination. The decontamination would remove a significant amount of
- radioactivity from the system and lower dose rates. The net benefit of lower dose
rates would amount to approximately 1000 person-rem saved during the
decommissioning of the facility. The actual DFs averaged approximately 15 and an
21
_ . _ _ . - - _ _.._ _.~____. _ . . _ . . _ _ . _ _ _ _ _ _ _ _ _ _ . . _
I
estimated 130 curies were removed from the system. The licensee estimates that
this will result in an overall dose savings of 900 person-rem during the
decommissioning.
c. Conclusions
The licenser measured and calculated the DFs for the RCS decontamination and
reported sw:essful results in lowering dose rates for future work.
- R1.2 Radicloaical Controls -
a. Insoection Scone (83750)
The inspector toured the radiologically controlled area (RCA) and discussed specific
radiological controls with radiation protection (RP) supervision and RP technicians.
The inspectors also reviewed radiological controls implemented for the RCS
decontamination effort including the radiation work permit (RWP), the Radiologica.
Safety Review 98-040, and associated radiological surveys.
l
b. ~ Observations and Findinos (
The inspectors toured various areas within the RCA. including the PAB, the !
containment building, and outside areas near the pipe trench openings. Appropriate
'
controls were observed for Radiation Areas and High Radiation Areas. Postings and
barriers were effectively placed to notify workers regarding changes in radiation
levels. Appropriate controls were noted to prevent the spread of radioactive i
contamination. The inspectors noted that workers were :aking the proper !
precautions for radiation protection as required by the licensee's staff. During
tours, good radiological housekeeping and good worker awareness of radiological
hazards was noted.
>
- No inadequacies were noted regarding the RWP or the Radiological Safety Review
98-040 for the RCS decontamination activities. The licensee had estimated that
4
RCS decontamination activities could be completed with an estimated 39 person- !
rem of exposure to workers. The actual totel dose for the decontamination of the
RCS was approximately 28 person-rem. Extensive use was made of teledosimetry
and cameras whose output were routed to several control points in low-dose areas.
This helped maintain exposures as low as is reasonably achievable (ALARA) by
. minimizing the need for routine operator staff rounds, routine RP coverage, and RP
surveys after each decontamination step. One good practice noted was that RP
staff irx omorated expected ranges of dose rates and contamination levels into the
Radiological Safety Review.
All workers in the containment building were assigned teledosimetry, in addition to
the electronic dosimeters and the thermoluminescent dosimeters used for routine
personnel monitoring, to allow real-time monitoring of their radiological exposure.
The units displayed the exposure results in the containment building control point,
22
- - . __. _- . _ . __ _ . _. _
the Health Physics desk on the operating floor, and at the control station in the
PAB. The system also allowed trending of exposures and exposure totals.
Additional controls were implemented for work in the pipe trench due to the highet
radiological contamination in some areas of the trench. For example, tents were
erected and portable high efficiency particulate air (HEPA) filtration units were
required with the exhaust directed to the plant ventilation system. All material from
the pipe trench was expected to be bagged and surveyed. Respirators and personal I
breathing zone analyzers (BZAs) were required for all workers making a PAB pipe
trench entry.
Up to the time of the inspection, whole body counting results of individuals
indicated that no workers had received any significant internal exposure during
chemical decontamination efforts. This was largely due to workers wearing
respirators. Additionally, there were no dose assignments made ts norkers from
skin or clothing contaminations.
c. Conclusions
Radiological controls for RCS decontamination work were well-planned and radiation
protection personnel maintained close oversight of work. Radiation protection was
effective in keeping workers' radiological exposures ALARA.
R2 Response to Events and Radiological Challenges
R2.1 Radioloaical Safety
a. Insoection Scooe (83750)
The inspectors reviewed the licensee's identification and corrective actions
associated with recent events or incidents, including actions taken in regard to the
spill of contaminated liquid to the PAB floor add pipe trench on July 27,1998, due
to the failure of valve LD-V-226.
b. Observations and Findinas
The spilled material / fluid released due to the failure of valve LD-V-226 contained
both radioactive and hazardous materials / fluids. Therefore, response to the spill
was initially limited to specially trained staff. Contamination surveys (smears)
indicated approximately 200,000 disintegrations per minute in a 100-centimeter
2
squared area (dpm/100cm )for beta and gamma emitting isotopes and 600
dpm/100cm' for alpha emitting isotopes. Significant potential airborne radioactivity
was expected if the liquid dried. The licensee attempted to keep the araa wet until
the clean-up could be performed.
Major clean-up efforts were conducted in chemical protective suits and respirators,
although the respkators were mainly used due to the potentially hazardous
chemicals. Potential effects of heat stress on workers were also monitored by
I
23
J
, .. .__ . _ ._ ._ . _ _ _ . . _ _ _ _ _ _ - . - . _ _ _ _ . -
lc )
i
I
i
i-
(
l licensee staff and stay times were provided to workers. Heat stress 'was of
L -
particular concern during pipe trench work activities because heat was transferred
from the elevated temperatures of the decontamination liquid in piping and there
was restricted air flow through the pipe trench. The inspectors noted that very .
good training was provided to first-line supervisors regarding heat stress prevention. !
However, some heat stress issues were still encountered during the work in the i
pipe trench. Licensee management responded to worker concerns by making ice : !
Jacket vests more readily available to those workers who desired to use them. j
1
The clean-up of the PAB was successfully completed within a few days. l
L
c. - Conclusions
Radiation Protection response to the spill from LD-V-226 was adequate, i
R2.2 Industrial Safety lasues
a. Inspection Scooe (71801) l
The purpose of this inspection was to review licensee actions in response to
industrial safety issues.
i
b. Observations and Findinas i
Severalissues occurred during the RCS decontamination that presented challenges I
to personnel safety.' NRC findings discussed in Sections 01 and M1 described
events involving the leakage of thermally hot and radioactivity contaminated water !
(leaks on July 27 and August 13), the loss of control of activated resin (August 11),
and the failure of the rigging on a 5 ton floor block (August 2).
Severalissues also occurred involving heat stress inside the containment and the
RHR pit. Conditions within these plant areas usually exhibit adverse temperature
and humidity conditions during the summer conditions. The containment
temperature also rose due to the gradual heat up of the reactor cavity to 140' F, I
which added heat and humidity to the containment. Containment area temperatures
rose to above 95u F. Similarly,in the RHR pit, the operating A RHR pump with
200' F fluid temperatures and no cooling to the RHR heat exchangers, caused the
area temperature to increase above 100o F, with local area temperatures greater
than 120' F.
Measures to help workers to deal with heat in the containment included the location
of a water cooler close to the containment access, the use of " cool tents" on the
charging floor, and the supply of ice vests. The licensee was slow to recognize the
deterioration in working conditions in the RHR pit, and bolstered heat stress controls
for the RHR pit only after two instances in which workers were almost overcome by
heat while working in the pit (reference ACRs98-629,656 and 661).
1-
'
24
. _ _ _ . --
_ - _ _ _ _ ___ _._ _ ._ .- . _ _ _ __ _._ ___._ _ _ . __
4
a c. Conclusions
The licensee was slow to respond to heat stress conditions in some plant areas.
Once focused on the issue, licensee controls of heat stress conditions improved.
.
- R2.3 Waste Water Manaaement
,
a .' insoection Scope (84750)
.
.
The purpose of this inspection was to review licensee actions to control RCS
decontamination waste water.
' b.
'
Observations and Findinas
During the RCS decontamination, plant events resulted in the leakage of
,
approximately 1200 gallons of decontamination process water on July 27,1998,
and approximately 125 gallons on August 13,1998. The water was contained
- . within the existing plant drain systems. The July 27,1998 event water was
' collected in sump (s) and was pumped into the Aerated Drain Tank (ADT) via sumps
in the RHR pit and the spent resin tank pit. The water that was released in the
August 13,1998 event was collected in the containment building sump and was
pumped to the ADT. The water from crains was collected in holding tanks until it
can be processed (reference drawings 16103 sheets 26018,26030,27049 and
50078). None of the waste water has been processed or discharged.
l The CORD process was designed to destroy the chemicals used during the
-
decontamination. However, the leaks occurred prior to the elimination of all the
chemicals. Although the licensee had planned to segregate the decontamination
.
' process water from normal olant process water, this segregation did not occur
because of the leaks. The waste water was controlled within the plant without
releases to the environment. The licensee and the State of Connecticut Department
of Environmental Protection (DEP) took samples of the affected tanks for analysis
and to characterize the waste.. The characterization results are expected to be
completed by mid-September,1998. The results of the evaluation will be used to
L
determine whether the waste water from decontamination activities'can be
processed under the current NPDES permit, or will require separate authorization
..
'
The consequence of this matter was that the licensee waste water tanks were near
- full capacity, such that the generation of additional waste water, such as from the
-
rain water collected within the radiologically controlled areas of the plant, further
challenged the control of tank inventories and building drains. The licensee planned
.
to develop a method to direct rain water from the outside diked area to the A waste
test tank for processing and release in accordance with the NPDES permit. This
matter remained under review at the conclusion of the inspection.
,
',
.
'
25
s
d
m...w.v.q- e .~
,+-s m - e. , .,.a - --
_ _ . . . _ - - _ . _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _
l
IV. Manaaement Meetinas
X1 Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management
l' periodically during the inspection, and during a meeting with Mr. R. Mellor and
!
Mr. J. Haseltine at the conclusion of the inspection on September 11,1998. The licensee
acknowledged the findings presented. . The inspector reviewed with the licensee whether
ar'v materials examined during the inspection should be considered proprietary. No
p. cAetary information was identified.
X3 Management Meetings
On August 3,1998, A.R. Blough, the Director of the Division of Nuclear Materials Safety,
l toured the site and together with other NRC Region I and NRR staff met with licensee
l
management on the status of decontamination activities. The NRC requested information
!
on root cause determinations, corrective actions, and readiness to proceed safely with the
next phase of the RCS decontamination. This meeting was open for public observation.
Following the conclusion of the meeting, NRC staff remained to answer questions from
members of the public.
L On August 20,1998, H. Miller, the NRC Region 1 Administrator, toured the site and met
with licensee management to discuss NRC concerns regarding licensee performance
highlighted by recent events. The NRC discussed the need to preclude operational events,
along with the improvements necessary in the areas of process controls, communications,
planning and preparation for future decommissioning activities and evolutions.
On September 3-4,1998, Mr. W. Axelson, Mr. T. Johnson, Mr. G. Pangburn,
Mr. T. Fredrichs and Mr. W. Raymond toured the site and conducted interviews with 30
Haddam Neck workers, supervisors and managers. The purpose of the reviews was to
obtain information on the organizational and management issues associated with the
l-
!
events during the RCS decontamination. The NRC observations were discussed in a
meeting with Mr. Dea Davis on September 4,1998.
While the RCS decontamination leaks and events were contained inside the plant and did
not endanger the public, the NRC remained concerned regarding the licensee's
performance. The NRC assessment to understand the reasons for the performance
problems continued at the end of this inspection period. The NRC findings to date indicate
i employees are not reluctant to raise concerns and that performance problems do not
appear to be linked to the licensee's process for maintaining a safety conscious work
environment.
l
26
!
I=
l
. , , - . . _ . , , - _ _ - . - - .
. . __ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . . . . _ _ _ _ _ _ _ . ~ _ _ _ _ . _ _ _
' PARTIAL LIST OF PERSONS CONTACTED
Russell Mellor, Vice President Operations and Decommissioning .
_
4
Gary Bouchard, Unit Director
.
. Kerry Harner, Chemistry Manager
Doug Heffernan, Maintenance Manager
Gerry Waig, Operations Manager
James Pandolfo, Security Manager -
Richard Sexton, Radiation Protection Manager
1 Gerry van Noordonnen, Nuclear Licensing
,
Pete Hollenbeck, Site Characterization Supervisor
'
' Keith Sickles, Design Engineer
'
. Edward Bingham, Engineering
John Haseltine, Engineering Director
Jay Tarzia, HP/ Chemistry Technical Support
William Symczack, RCS Decontamination Program Manager
Robert Sojka, CYAPCO Decentamination Manager
Bert Mayer, Decontamination Project System Engineer
"
Frank Gilbert, Decontamination Project System Engineer
Patrick Holmes, Decontamination Project System Engineer
. Robert Pritchard, Engineering Supervisor, Mechanical Systems
INSPECTION PROCEDURES USED
-
IP 37700: ' Design Changes and Modifications
IP 37801: Safety Reviews, Design Changes, and Modifications at PSRs !
IP 40801: Self-Assessment, Auditing, and Corrective Action
IP 42700: Plant Procedures
4
IP 61726: Surveillance Observations
IP 62707: ' Maintenance Observations
^
- IP 62801: j
Maintenance and Surveillance at Permanently Shutdown Reactors
IP 71707: ' Operational Safety Verification {
-!
- IP 71801:1 Decommissioning Performance and Status Review at PSRs
!
- IP 83750: Occupational Radiation Exposure
IP_84750: RadWaste Treatment, and Effluent & Environmental Monitoring
IP 92702: Follow-up on Corrective Actions for Violations and Deviations ;
IP 93702: Prompt Onsite Response to Events at Operating Power Reactors i
>
ITEMS OPEN, CLOSED, AND DISCUSSED
DEAD -
_
'
98-04-01 VIO Inadequate Procedure for RCS Decontamination (Leak Check)
98-04-02 VIO Inadequate Procedure for RCS Decontamination (Pressure Control) l
i
98-04-03 VIO inadequate Procedure for Purification Operation (Flows)
98-04-04 URI Actione to Address Weaknesses in Rigging Program
,
'
98-04-05 VIO Inadequate Procedure for Rigging Floor Blocks
27
, . , - . - . . . . , - .
- _-. _-. ._ . . _ . _ . _ _ _ . . _ . _ _ _ _ _.____ _ _ _ _ . . . _ _
_ . _ . . _ _ _ . _ . _ -
l
.-
l
LIST OF ACRONYMS USED
!. ACP Administrative Control Procedure
l ACR Adverse Condition Report
[ ADT Aerated Drain Tank
.
'AEOD
_
.
l Office for Analysis and Evaluation of Operational Data
l ALARA '
As Low As is Reasonably Achievable
{ ' AOP . Abnormal Operating Procedure
!
BZA . Breathing Zone Analyzers i
!
-CCTV Closed Circuit Television !
'CFR Code of Federal Regulations
CVCS' Chemical and Volume Control System !
'
CYAPCo Connecticut Yankee Atomic Power Company
DEP Department of Environmental Protection
DF- Decontamination Factor
F' Fahrenheit ~
gpm gallons per m;aute
i HEPA high efficiency particulate air
HP Health Physics t
HPSI' High Pressure Safety injection I
L ips - inches per second
IR inspection Report
LDBP Letdown Booster Pump
LDRP Letdown Return Pump
MDP Master Decontamination Procedure
-NOP Normal Operating Procedure
NOV Notice of Violation
NRC Nuclear Regulatory Commission
PAB Primary Auxiliary Building
PDR Public Document Room -
L RCA: Radiological Controlled Area
RP: Radiation Protection
RWPs Radiation Work Permits
RWST Refueling Water Storage Tank
SAR Safety Analysis Report
SFP Spent Fuel Pool
UFSAR Updated Final Safety Analysis Report
-VCT. Volume Control Tank
WCM Work Control Manual
l
l
l
l
28
.
__ . _ . . _ _ _ _ _ _ . . . . . . __ . - . . . _ . - _ _ _ . _ _ . _ . _ _ . _ __ _ . _ . _ . _ _ _ . _ . . _ . .
l
TABLE 1
RCS DECONTAMINATION
MAJOR MILESTONES
DATE(s) EVENT
July 17-18 Additional verification of decontamination valve boundary
July 21 Begin RCS decontamination; A RHR and letdown pumps in service
l
July 22 Start system heat up to 200 F
July 23-26 Hold for purification system alignments and repairs
l
July 26 10:00 p.m. - start Cycle 1 permanganate acid injection
July 27 10:30 a.m. - Cycle 1 oxalic acid injection
July 27 12:38 p.m. - 1200 gallon leak of RCS decontamination fluid
July 27 - Leak recovery, repairs, event investigation and corrective actions
August 10
August 10 10:01 a.m. - Start Cycle 2 injection of oxalic acid
August 11 3:35 a.m. - C Domineralizer/Postfilter failure - loss of resin
August 13 Licensee decision to end decontamination after Cycle 2
August 13 7:07 p.m. - LDBP discharge pipe failure - 125 gallon leak
August 13-18 Leak recovery and repairs
August 16 Recommence final cleanup for Cycle 2
August 22 Alternate RCS cleanup using RHR purification !
August 22 Start RCS cooldown via RHR cooling
August 24 Cycle 2 cleanup complete
August 24 -
September 11 A RHR remains in service for cooling and further RCS cleanup
4
29
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V
FIGURE 1 - CONNECTICUT YANKEE 'i
RCS FULL SYSTEM DECONTAMINATION FLOW PATH
. - .
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Attachment I
Status of Reactor Coolant i
System Decontamination.
August 3,1998
'
Connecticut Yankee Atomic -
Power Company
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Agenda
Introc!uction - Russ Mellor -
,i
CY Decontamination ::
! Process - Bob Sojka
Causes - John Haseltine
!
Corrective Actions - Noah Fetherston
Summary - Russ Mellor
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CY Decontamination Process
Bob Sojka
Decontamination Manager
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- CY Decontamination
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Planned for One Year
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Includes Lessons Learned From Others
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Utilizes Industry Experts
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Termination Criteria Specified
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Using 22 Remote Cameras and 110
l
1
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Remote Radiation Monitors
- 20 Detailed Procedures Developed
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- Plant Installed Equipment inside Buildings -
- Extensive Evaluation of Diaphragm Valves j
.
_ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
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Summary of RCS Chemical
Decontamination Process at CY
'
3 to 5 Cycles
2 to 5 Days per Cycle j
- > 50% Activity Removal per Cycle
a
- Overall Reduction of 100C rem for
Decommissioning Planned
Contingency Plans in Place
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Cycle Description
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Heat system to 200 degrees F
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Add Permanganic Acid to oxidize crud layer
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Add Oxalic Acid to Destroy Permanganic Acid and
Complex Metals
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Flow Through Plant Demineralizer Systems for Clean-Up
~
(Cation)
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Once " Clean" Fluid Exposed to High Intensity UV Light -
To Decompose Acid - Results in Water and CO 2
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Prepare for Next Cycle
Manage Resin Beds
Survey for Radiation
.
. _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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. Event Description
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System Status Prior to Transient
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Transient Condition
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Staff Response and Actions
Leak Isolation
,
Notification to State and NRC Officials ,
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Event impacts
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Leaks Were Contained within Plant a
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- -
No Release to the Environment -
.
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No Contamination of Employees
Extent of Leak
'
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Event Response Consistent With
Contingency Plan
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Event Team Formed
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Team Comprised of Members from
Oversight, Operations, Engineering and
Corrective Actions .
Determine Cause
Identify Contributing Factors
Develop Recommendations
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Contributing Factors
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Repeated Attempts to Line-up Demineralizer without ,!
Changing Condition i
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- Influenced by Carbon Dioxide Entrapment ;
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- Time Pressure to Strip out Radionuclides
Focus on Letdown Flow Rate and VCT Level
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Magnitude of Pipe Vibration not Effectively
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Communicated to Control Room ,
-
Lack of Guidance and Action for Potential Pressure
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Contributing Factors
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Human Factor / Material Condition of Reach Rod,
Piping and Valves
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Corrective Actions
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Repairs and Modifications
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Leak Repairs .
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Safety Valve Replacement and Testing
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Auto Pump Trip Circuit
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System Inspections and Testing
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Structural Integrity inspection
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Pressure Tests
Leak Inspections (Elevated Pressures) !;
Flow Testing ;
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Corrective Actions
Procedure Changes
'
e improved Control of Hydraulic Characterics
Limiting System Flow Rates during Decon
'
Termination Criteria - Pressure and Flow
Analyze for Carbon Dioxide Buildup during Decon
Lessons Learned Briefings with Operating
Crews
Operation of Valves Require " Hand Turn
Counts" during Decon
_ - -___ -__ _ -______ _____ - __________ _ _ _ _ - - _ _ _ _ - - - 1
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Summary
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Experienced a Pressure Transient Event '
and Leak
Operator Actions Responding to Leak
Were Excellent .!
No Worker or Environmental Impact
Corrective Actions Address Any Potential
Overpressurization
l
Summary
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i!
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Repairs and Physical Enhancements L
Complete and System Testing
l
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Enhancements to Procedure implemented
l
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> 50% Dose Reduction Achieved During
j . Cycle One 3
- a
i
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Ready to Complete Decontamination when :
- all Corrective Actions are Completed !
!
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{
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Attachment 11
References and Documents Reviewed
1.
CYDE 98-644, Lessons Learned ACR 98-620,8/8/98
2.
3. Root Cause Report, ACR 98-620: Mechanical Failure in Purification
4. Root Cause Report, Mispositioned SI Valves for Tag Out 98020
5. Root Cause Report, Mechanical Failure in Purification System Revision 2
6. NCR 98-0073, Liquid Penetrant Exam of 1" dia. LD8P Recirc. Line Welds
7. NCR 98-0072, LDBP Recirc Line Visual and LPE of Sock-O-Let and Welds
SPL 10.11-1, Master Decontamination Procedure
8.
9. SPL 10.11-9, Leak Check Of Purification Demineralizer and Piping
10. SPL 10.11-30, Leak Check Of Purification Demineralizer 1-
11. SPL 10.11-24, Preoperational Flow Test and Leak Check
12. MMOD CY 98507, Jumper Reactor Pressure Vessel Head Deluge to Pressurizer
13. MMOD CY 98508 and 98509, Letdown Return Pump
14. MMOD CY 98511, Spent Fuel Pool lon Exchanger Return Jumper
15. MMOD CY 98512, Letdown Orifice Removal and Piping Replacement
17. AWO CY 98-02289,2"-AC-151R-101 Temporary Repair
18. AWO CY 98-02278, Removal of Shield Blocks to Access LD-RV-252
19. AWO CY 98-02230, Letdown Booster Pump High Pressure Cutout Switch
20. Proto-Power Calculation 98-012, Decontamination System Thermal-Hydraulic
21. CY-98-136, Reactor Coolant System Decontamination
22. CYDE 98-0222, RCS Decon Piping StructuralIntegrity Assessment
23. CY-JDH-98-025, Proto Power Model Testing for LD-V-238
24. ERT Report, Mispositioned Pdmary Water isolation Valve
25. Preliminary Analyris of LDV266 Stub Tube Failure, dated August 7,1998
26. Metallurgical Analysis Report - DE&S Examination of Cracked Pipe / Valve LD-V-
27. Metallurgy Sample #2391 - Preliminary Assessment of Pipe at Valve LD-V-226
28. P&lD 16103-26018Sh.1 - Chemical & Volume Control Letdown to VCT
29. P&lD 16103-26018 Sh. 2 - Chemical & Volume Control Purification
30. P&lD 16103-26028- Residual Heat Removal System
31. P&lD 16103-26030Sh. 5 - Liquid Waste System Spent Resin Building
32. P&lD 16103 26030Sh. 4 - Liquid Waste System Primary Drain Header
33. P&lD 16103 - 26030 Sh. 2 - Liquid Waste System Aerated Drains Tanks
34. P&lD 16103-27049 Arrangement lon Exchangers and Resin Storage
36. P&lD 16103-26018 Sh. 2&6 - Chemical & Volume Con
37. DPM 98-155, Readiness to Proceed with Decontamination Cycle 2 & 3
38. CY-GHB-98-134, Apparent Cause Report, ACR 98646- Dropped PA8 Floor Block
39. AWO 98-2444, SFP Filter inspection
40. AWOs 97-4324 and 98- 2119,2134, Postfilter FL-65-1 A inspection and Cleaning
41. Jumper Device 98-36, Postfilter FL-65-1 A Element Replacement
42. Technical and Safety Evaluation DP 98-145,IS-1-1C Postfilter Retention Element
43. SPL 10.1-48, RHR Purification System Operation
44. NOP 2.7-1, Reactor Coolant Letdown Post Filter and Purification Demineralizers
45. SPL 10.1-54, Alternate RHR Purification
f
i
,
_ - - -
_ _ _ . _ __ . _ _ _ . . _ . _ _ _ . _ . _ _ . _ _ _ . _ . _ . _ _ _ . _ _ _
i
Attachment 11
References and Documents Reviewed
-
1.
CYDE 98-644, Lessons Learned ACR 98-620,8/8/98
2.
3. Root Cause Report, ACR 98-620:MechanicalFailure in Purification
4. Root Cause Report, Mispositioned St Valves for Tag Out 98020
5. Root Cause Report, Mechanical Failure in Purification System Revision 2
6. NCR 98-0073, Liquid Penetrant Exam of 1" dia. LDBP Recire. Line Welds
7. NCR 98-0072,LDBP Recirc Line Visual and LPE of Seck-O Let and Welds
SPL 10.11-1, Master Decontamination Procedure
8.
9. SPL 10.11-9, Leak Check Of Purification Demineralizer and Piping
SPL 10.11-10, Flush of SFP lon Exchanger and HPSI Header Discharge Piping
10. SPL 10.11-30, Leak Check Of Purification Domineralizer 1-1-1B
11. SPL 10.11-24, Preoperational Flow Test and Leak Check
12. MMOD CY 98507, Jumper Reactor Pressure Vessel Head Deluge to Pressurizer
13. MMOD CY 98508 and 98509, Letdown Return Pump
14. MMOD CY 98511, Spent Fuel Pool lon Exchanger Return Jumper
15. MMOD CY 98512, Letdown Orifice Removal and Piping Replacement
16. AWO CY 98-02226, Letdown Header Drain isolation Temporary Repair Jumper
17. AWO CY 98-02289,2"-AC-151R-101 Temporary Repair / Jumper
18. AWO CY 98-02278, Removal of Shield Blocks to Access LD-RV-252
19. AWO CY 98-02230, Letdown Booster Pump High Pressure Cutout Switch
20. Proto-Power Calculation 98-012, Decontamination System Thermal-Hydraulic Model
21. CY-98-136, Reactor Coolant System Decontamiriation
22. CYDE 98-0222, RCS Decon Piping Structural Integrity Assessment
23. CY-JDH-98-025, Proto Power Model Testing for LD-V-238
24. ERT Report, Mispositioned Primary Water Isolation Valve
25. Preliminary Analysis of LDV266 Stub Tube Failure, dated August 7,1998
26. Metallurgical Analysis Report - DE&S Examination of Cracked Pipe / Valve LD-V-226
27. Metallurgy Sample #2391 - Preliminary Assessment of Pipe at Valve LD-V-226
28. P&lD 16103-26018Sh.1 - Chemical & Volume Control Letdown to VCT
29. P&lD 16103-26018Sh. 2 - Chemical & Volume Control Purification
30.'P&lD 16103-26028- Residual Heat Removal System
31. P&lD 16103-26030Sh. 5 - Liquid Waste System Spent Resin Building
32. P&lD 16103-26030Sh. 4 - Liquid Waste System Primary Drain Header
33. P&lD 16103 - 26030 Sh. 2 - Liquid Waste System Aerated Drains Tanks
34. P&lD 16103-27049 Arrangement lon Exchangers and Resin Storage
35. P&lD 10899-FC-33A& B Dets. Sh.1& 2 lon Exchangers and Resin Storage Building
-
36. P&lD 16103 26018 Sh. 2&6 - Chemical & Volume Control (C&VC) Purification
37. DPM 98-155, Readiness to Proceed with Decontamination Cycle 2 & 3
38. CY-GHB-98-134, Apparent Cause Report, ACR 98646- Dropped PAB Floor Block
39. AWO 98-2444, SFP Filter inspection
40. AWOs 97-4324 and 98- 2119,2134, Postfilter FL-65-1 A Inspection and Cleaning
41. Jumper Device 98-36, Postfilter FL-65-1 A Element Replacement
j
42. Technical and Safety Evaluation DP 98-145,IS-1-1C Postfilter Retention Element
43. SPL 10.1-48, RHR Purification System Operation
44. NOP 2.7-1, Reactor Coolant Letdown Post Filter and Purification Demineralizers
45. SPL 10.1-54, Alternate RHR Purification
. . . . , . - _ _ - _ _ . . _ _ __- - - - - l
. .- - - - .- . - - . . . . -.- - - . - - - - - . - -.- - . -.. .
I
{ Attachment ll
l References and Documents Reviewed
f
!
1. CYDE 98-644, Lessons Learned ACR 98-620,8/8/98
2. Root Cause Report, ACR 98-620:MechanicalFailure in Purification
c 3. Root Cause Report, Mispositioned SI Valves for Tag Out 98020
l 4. Root Cause Report, Mechanical Failure in Purification System Revision 2 )
5. NCR 98-0073, Liquid Penetrant Exam of 1" dia. LDBP Recire. Line Welds !
6. NCR 98-0072, LDBP Recirc Line Visual and LPE of Sock-O-Let and Welds
7. SPL 10.11-1, Master Decontamination Procedure
8. SPL 10.11-9, Leak Check Of Purification Demineralizer and Piping
9.
SPL 10.11-10, Flush of SFP lon Exchanger and HPSI Header Discharge Piping
10. SPL 10.11-30, Leak Check Of Purification Demineralizer 1-1-1B
11. SPL 10.11-24, Preoperational Flow Test and Leak Check !
l 12. MMOD CY 98507, Jumper Reactor Pressure Vessel Head Deluge to Pressurizer
!
13. MMOD CY 98508 and 98509, Letdown Return Pump
i
1 14. MMOD CY 98511, Spent Fuel Pool lon Exchanger Return Jumper
1
!
15. MMOD CY 98512, Letdown Orifice Removal and Piping Replacement
16. AWO CY 98-02226, Letdown Header Drain isolation Temporary Repair Jumper !
'
17. AWO CY 98-02289,2"-AC-151R-101 Temporary Repair / Jumper
,
18. AWO CY 98-02278, Removal of Shield Blocks to Access LD-RV-252
19. AWO CY 98-02230, Letdown Booster Pump High Pressure Cutout Switch
)
20. Proto-Power Calculation 98-012, Decontamination System Thermal Hydraulic Model '
21. CY-98-136, Reactor Coolant System Decontamination
l 22. CYDE 98-0222, RCS Decon Piping Structuralintegrity Assessment
! 23. CY-JDH-98 025, Proto Power Model Testing for LD-V-238
24. ERT Report, Mispositioned Primary Water isolation Valve !
25. Preliminary Analysis of LDV266 Stub Tube Failure, dated August 7,1998
26. Metallurgical Analysis Report - DE&S Examination of Cracked Pipe / Valve LD V-226
l- 27. Metallurgy Sample #2391 - Preliminary Assessment of Pipe at Valve LD-V-226
l
28. P&lD 16103-26018Sh.1 - Chemical & Volume Control Letdown to VCT
29. P&lD 16103-26018Sh. 2 - Chemical & Volume Control Purification i
30. P&lD 16103-26028- Residual Heat Removal System
! 31. P&lD 16103-26030Sh. 5 - Liquid Waste System Spent Resin Building
32. P&lD 16103-26030Sh. 4 - Liquid Waste System Primary Drain Header
33. P&lD 16103 - 26030 Sh. 2 - Liquid Waste System Aerated Drains Tanks
34. P&lD 16103-27049 Arrangement lon Exchangers and Resin Storage
l
' 35. P&lD 10899-FC-33A& B Dets. Sh.1& 2 lon Exchangers and Resin Storage Building
36. P&lD 16103-26018 Sh. 2&6 - Chemical & Volume Control (C&VC) Purification
37. DPM 98-155, Readiness to Proceed with Decontamination Cycle 2 & 3
i 38. CY-GHB-98-134, Apparent Cause Report, ACR 98646- Dropped PAB Floor Block
l 39. AWO 98-2444, SFP Filter inspection
40. AWOs 97-4324 and 98- 2119,2134, Postfilter FL-65-1 A Inspection and Cleaning
41. Jumper Device 98-36, Postfilter FL-65-1 A Element Replacement
42. Technical and Safety Evaluation DP 98-145,IS-1-1C Postfilter Retention Element
l 43. SPL 10.1-48, RHR Purification System Operation
- 44. NOP 2.7-1, Reactor Coolant Letdown Post Filter and Purification Demineralizers
l 45. SPL 10.1-54, Alternate RHR Purification
I
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