ML20138J819
| ML20138J819 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/17/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138J812 | List: |
| References | |
| 50-302-97-06, 50-302-97-6, NUDOCS 9705090047 | |
| Download: ML20138J819 (9) | |
See also: IR 05000302/1997006
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U.S. NUCLEAR REGULATORY COMMISSION
REGION 2
Docket No:
50-302
License No:
Report No:
50-302/97-06
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Licensee:
Florida Power Corporation
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Facility:
Crystal River 3 Nuclear Station
Location:
15760 West Power Line Street
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Crystal River, FL 34428-6708
Dates:
January 27 through March 21,1997
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Inspector.
R. Schin, Reactor Inspector
Approved by:
H. Christensen, Chief, Engineering Branch
Division of Reactor Safety
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Enclosure
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9705090047 970417
ADOCK 05000302
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EXECUTIVE SUMMARY
Crystal River 3 Nuclear Station
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NRC Inspection Report 50-302/97-06
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This specialinspection addressed the engineering functional area. The purpose of the
inspection was to follow up on Unresolved item (URI) 97-01-06; HPl System Design,
Licensing Basis, and TS Concems.
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Enaineerina
An apparent Violation (eel 50-302/97-06-01) was identified for inadequate safety evaluations
for added operator actions for design basis small break loss of coolant accident (SBLOCA)
mitigation. The 10 CFR 50.59 safety evaluation for Final Safety Analysis Report (FSAR)
Rev. 23 was inadequate in that it did not address the fact that the increase in required
operator actions for design basis SBLOCA mitigation would result in both the possibility of a
malfunction of a different type (operator error) and an increase in the probability of
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occurrence of a malfunction of equipment important to safety. The safety evaluation did not
recognize that, per 10 CFR 50.59, prior NRC review and approval was required. The safety
evaluation was also inadequate in that it did not address two of the seven required operator
actions in the new licensee SBLOCA calculation or assure that they were included in the
FSAR. In addition, four 50.59 safety evaluations [for FSAR Rev. 23; Short Term Instruction
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(STI) 95-0061; STI 96-008; and Emergency Operating Procedure EOP-03, Rev. 4] were
inadequate in that they failed to address the potentialincrease in the probability of Reactor
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Coolant Pump (RCP) seal failure resulting from isolating RCP seal injection.
The inspectors assessed the licensee's performance concerning the five areas c.. continuing
NRC concern described in the following paragraph. The assessment is limited to the specific
issue addressed in the respective paragraph.
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NRC AREA OF CONCERN
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ASSESSMENT PARAGRAPH
E8.1
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Management Oversight
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Engineering Effectiveness
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Knowledge of Design Basis
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Compliance With Regulations
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Operator Performance
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S = Superior G = Good A = Adequate / Acceptable I = Inadequate
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Blank = Not Evaluated / Insufficient Information
E8.1: URI 50-302/97-01-06; HPI System Design, Licensing Basis, and TS Concerns
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Report Details
L Enaineerina
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Miscellaneous Engineering issues
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E8.1
(Open) URI 50-302/97-01-06: HPl System Desian. Licensina Basis. and TS Concems
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(Closed) LER 50-302/96-006-01: Consideration of instrument Error Results in
Unacceptable Marain for HPi Flow in SBLOCA Analysis
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Inspection Scope (92903)
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During this inspection, the inspector followed up on the first two (of three) exampies of
this unresolved item (URI).
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1)
The first example of this URI was a concern about the number of required
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operator actions in the licensee's design to mitigate a design basis small break
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loss of coolant accident (SBLOCA). The seven required operator actions in
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the licensee's current calculation for small break loss of coolant accidents
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(LOCAs); M96-0032, Reevaluation of HPl Requirements During Small Break
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LOCAs, dated May 2,1996; were more than the one required operator action
approved by the NRC in the 1979 licensing basis. The seven were also more
than the five required operator actions in Final Safety Analysis Report (FSAR)
Rev. 23, dated November 18,1996, and more than the two required operator
actions in the previous FSAR revision. The inspector had noted that the
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increase in required operator actions could cause an increase in the probability
of occurrence of a malfunction of equipment important to safety and the
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possibility of a malfunction of a different type (operator error) and therefore,
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per 10 CFR 50.59, prior NRC review and approval could be required.
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However, the inspector had considered that further review of this issue was
needed to determine if any of the added operator actions had been reviewed
and approved by the NRC between 1979 and 1996. In addition, inspector
review was needed of any procedure, plant design, or FSAR changes (and
related 50.59 safety evaluations) that added required operator actions or
deleted automatic actions.
2)
The second example of this URI was a concern with statements made in
Calculation M96-0032 and in a licensee letter to the NRC dated May 22,1996,
titled "New Small Break Loss-of-Coolant Accident (SBLOCA) Analyses." The
statements of concern indicated that previous SBLOCA analyses had
incorrectly assumed that reactor coolant pump (RCP) seal injection and normal
makeup were automatically isolated, based on a generic Babcock and Wilcox
(B&W) plant. The licensee's SBLOCA calculations had been performed by
Framatome Technologies incorporated (FTI), which was formerly B&W. The
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inspector had considered that further review was needed of the potential
design control error, its impact on past operability, related reportability, and
generic applicability to other B&W plants.
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Qbservations and Findinas
1)
The seven operator actions that were currently required by the licensee, per
Calculation M96-0032, to mitigate the spectrum of design basis small break
LOCAs included:
(1)
trip all running RCPs within two minutes,
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initiate high pressure injection (HPI) flow through all four injection lines
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within 10 minutes,
(3)
isolate letdown within 10 minutes,
(4)
isolate RCP sealinjection within 20 minutes,
(5)
isolate normal makeup within 20 minutes,
(6)
ensure adequate HPl flow (isolate a broken injection line) within 20
minutes, and
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(7)
ensure adequate emergency feedwater (EFW) flow within 20 minutes.
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Of those, FSAR Rev. 23 included operator actions 2,3,4,5, and 6. It did not
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include operator actions 1 or 7. The previous FSAR revision included operator
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action 2 above and an action to balance flows in the HPI injection lines.
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The inspector noted that the emergency core cooling system (ECCS) and other
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plant safety systems were initially designed to operate essentially automatically
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for the first 20 minutes of a design basis event. One operator action was
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approved by the NRC as part of the CR-3 design for SBLOCA mitigation; in an
NRC Safety Evaluation dated May 29,1979; and that was operator action 2
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above. During this inspection, neither the inspector nor the licensee identified
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any NRC approval of licensee design reliance on the other six operator actions
listed above.
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The inspector reviewed the 10 CFR 50.59 safety evaluation for FSAR Rev. 23,
dated April 30,1996, titled "FSAR Revision due to HPl Reevaluation." The
purpose of this FSAR revision was to incorporate the results of the licensee's
new calculation for design basis small break LOCAs; M96-0032, Reevaluation
of HPl Requirements During Small Break LOCAs, dated May 2,1996. There
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was no separate safety evaluation for the calculation. The safety evaluation
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for FSAR Rev. 23 did not address the fact that the increase in required
operator actions would result in both the possibility of a malfunction of a
different type (operator error) and an increase in the probability of occurrence
of a malfunction of equipment important to safety. Also, the safety evaluation
did not recognize that, per 10 CFR 50.59, prior NRC review and approval was
required. Therefore, the inspector concluded that the 10 CFR 50.59 safety
evaluation for FSAR Rev. 23 was inadequate. This issue is identified as the
first example of Escalated Enforcement item (EEI) 50-302/97-06-01,
Inadequate Safety Evaluations for Added Operator Actions for Design Basis
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SBLOCA Mitigation.
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The safety evaluation for FSAR Rev. 23 also did not address the fact that two
of the seven required operator actions in the new calculation were not included
in the FSAR. Also, there was no separate safety evaluation for Calculation
M96-0032. The failure of the safety evaluation for FSAR Rev. 23 to address
two of the seven required operator actions in Calculation M96-0032, and to
assure that they were added to the FSAR, is identified as the second example
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of eel 50-302/97-06-01, Inadequate Safety Evaluations for Added Operator
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Actions for Design Basis SBLOCA Mitigation.
Licensee and inspector review of emergency operating procedures (EOPs)
determined that all but one of the seven required operator actions had been in
the EOPs since the 1974 - 1979 time period. The inspector noted that the
EOPs typically included many operator actions that were outside of the design
basis as well as some that were relied upon in the design basis. Also, the
licensee and inspector determined that there had been no plant modifications
that deleted automatic actions assumed by the NRC Safety Evaluation dated
May 29,1979.
Operator action 4 above (isolate RCP sealinjection) was initially added to the
EOPs by STl 95-0061, which was effective from November 8,1995 to
February 8,1996. It was again temporarily added to the EOPs by STI 96-008,
which was effective from February 8,1996, to May 6,1996. This operator
action was then permanently added to the EOPs by EOP-03, Rev. 4, dated
May 2,1996. Both of these STls and EOP-03, Rev. 4 had been reviewed by
the Plant Review Committee (PRC) and approved by the Director, Nuclear
Plant Operations (DNPO). The inspector reviewed the 10 CFR 50.59 safety
evaluations for each of these three procedure changes and for FSAR Rev. 23,
and noted that none of them addressed the potential consequent increase in
the probability of RCP seal failure. The inspector noted that the isolation of
RCP sealinjection could cause an increase in the RCP seal temperature and a
change in the seating of the seal. Subsequent initiation of RCP sealinjection
could cause a decrease in RCP seal temperature and a change in the seating
of the seal. The inspector further noted that changes in RCP sealinjection at
other sites, without appropriate precautions and controls, have in the past
resulted in RCP seal failures. Consequently, the inspector concluded that the
four 10 CFR 50.59 safety evaluations were inadequate in that they failed to
address the potentialincrease in the probability of RCP seal failure. These
inadequate safety evaluations for an added operator action are a third example
of eel 50-302/97-06-01, Inadequate Safety Evaluations for Added Operator
Actions for Design Basis SBLOCA Mitigation.
2)
During this inspection, the inspector reviewed the statements of concern in
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Calculation M96-0032 and in the licensee letter to the NRC dated May 22,
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1996, titled "New Small Break Loss-of-Coolant Accident (SBLOCA) Analyses;"
reviewed Problem Reports96-018 and 96-035 and related LER 50-302/96-006-
01, titled " Consideration of Instrument Error Results in Unacceptable Margin for
HPl Flow in SBLOCA Analysis," and discussed them with licensee and FTl
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engineers. The inspector also reviewed the FTl HPI flow assumptions for the
SBLOCA calculations, which were in a separate proprietary document. Based
on this review, the inspector concluded that the statements of concern in
Calculation M96-0032. The SBLOCA calculations assumed a certain HPl
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injection flow (for Crystal River 3 and other B&W plants), but did not assume
any particular system alignment to get that flow and also did not account for
instrument errors or uncertainties. To rely on the SBLOCA calculations, each
licensee had to assure that their HPl injection flow was equal to or greater than
that assumed in the calculations. Crystal River's prior analyses, operating
procedures, and surveillance testing procedures had appeared to do that in the
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past. EOPs had operators balance the HPl injection Sws to specified values
(above the SBLOCA calculation flow requirements and below HPI pump runout
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flowrates). However, new instrument error calculations resulted in a larger
allowance for error for the HPI flow instruments which in turn resulted in a
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licensee conclusion in January 1996 that the previous HPl injection flows had
not assured the SBLOCA calculation flow requirements would be met (the flow
deficit was a few gpm). Since the licensee could not increase HPI pump
flowrates due to potential pump runout, they elected to isolate RCP seal
injection flow (approximately eight gpm per RCP, or 32 gpm total) to obtain the
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needed increase in HPl injection flow to the reactor. The FTl engineers stated
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that lack of HPl flow margin was not a generic B&W plant problem - they
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described how other B&W plants had different HPl piping arrangements and
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did not have the problem. The inspector concluded that, other than the
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licensee's allowance for HPl injection flow instrument error, there was no
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identified design control error with the HPI flow assumptions for the SBLOCA
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analyses. The inspector also concluded that the licensee had adequately
reported the issue in LER 96-006-01. The inspector verified that all but one of
the licensee's corrective actions in the LER had been completed, including a
revised SBLOCA analysis and an EOP revision. One action that was not
completed was consideration of plant modifications to eliminate the need for
the manual operator action to isolate RCP seal injection, but that was in
progress. However, the inspector concluded that the licensee's safety
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evaluations for the increased operator actions did not meet NRC requirements,
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as discussed above. LER 96-006-01 is closed.
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c.
Conclusions
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The inspector concluded that the 10 CFR 50.59 safety evaluation for FSAR Rev. 23
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was inadequate in that it did not address the fact that the increase in required
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operator actions for design basis SBLOCA mitigation would result in both the
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possibility of a malfunction of a different type (operator error) and an increase in the
probability of occurrence of a malfunction of equipment important to safety. The
safety evaluation did not recognize that, per 10 CFR 50.59, prior NRC review and
approval was required. The safety evaluation was also inadequate in that it did not
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address two of the seven required operator actions in the new licensee SBLOCA
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calculation or assure that they were included in the FSAR. In addition, four 10 CFR
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50.59 safety evaluations (for FSAR Rev. 23; STI 95-0061; STI 96-008; and EOP-03,
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Rev. 4) were inadequate in that they failed to address the potential increase in the
probability of RCP seal failure resulting from isolating RCP seat injection. This issue
is identified as eel 50-302/97-06-01, inadequate Safety Evaluations for Added
Operator Actions for Design Basis SBLOCA Mitigation,
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The inspector assessed the licensee's performance, with respect to this issue, in the
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five areas of continuing NRC concern:
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Management Oversight - Inadequate
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Engineering Effectiveness - Inadequate
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Knowledge of the Design Basis -Inadequate
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Compliance with Regulations - Inadequate
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Operator Performance - Not Applicable
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jlh Manaaement Meetinas
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Exit Meeting Summary
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The inspection scope and findings were summarized in an exit meeting held on
March 21,1997. Proprietary information is not contained in this report. Dissenting
comments were not received from the licensee.
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PARTIAL LIST OF PERSONS CONTACTED
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Licensees
R. Anderson, Senior Vice President, Nuclear Operations
J. Baumstark, Director, Quality Programs
G. Becker, Manager, Operations
J. Campbell, Assistant Plant Director, Maintenance
K. Campbell, Senior Nuclear Operations Engineer
W. Conklin, Jr., Director, Nuclear Operations Materials and Controls
J. Cowan, Vice President, Nuclear Production
R. Davis, Assistant Plant Director, Operations
B. Gutherman, Manager, Nuclear Licensing
G. Halnon, Assistant Director, Nuclear Operations Site Support
B. Hickle, Director, Nuclear Plant Operations
J. Holden, Director, Nuclear Engineering and Projects
D. Kunsemiller, Director, Nuclear Operations Site Support
NRC
C. Christensen, Chief, Engineering Branch, Division of Reactor Safety
J. Jaudon, Director, Division of Reactor Safety
K. Landis, Chief, Reactor Projects Branch 3, Division of Reactor Projects
S. Cahill, Senior Resident inspector
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T. Cooper, Resident inspector
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INSPECTION PROCEDURES USED
IP 92903:
Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Type item Number
Status Description and Reference
50-302/97-06-01
Open
inadequate Safety Evaluations for Added
Operator Actions for Design Basis
SBLOCA Mitigation. (paragraph E8.1)
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Closed
Tvoe item Number
Status
Description and Reference
LER
50-302/96-006-01
Closed
Consideration of Instrument Error Results
in Unacceptable Margin for HPl Flow in
SBLOCA Analysis. (paragraph E8.1)
Discussed
Tvoe item Number
Status
Description and Reference
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50-302/97-01-06
Open
HPl System Design, Licensing Basis, and
TS Concems. (paragraph E8.1)
LIST OF ACRONYMS USED
B&W - Babcock and Wilcox
CR-3
- Crystal River Unit 3
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DNPO - Director, Nuclear Plant Operations
- Enforcement Action
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ECCS - Emergency Core Cooling System
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- Escalation Enforcement item
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- Emergency Feedwater
- Emergency Operating Procedure
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- Engineerni Safeguards
- Florida Pcwer Corporation
FSAR - Final Safety Evaluation Report
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FTl
- Framatome Technologies incorporated
"PI
- High Pressure Injection
LER
- Licensee Event Report
LOCA - Loss of Coolant Accident
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- Plant Review Committee
- Reactor Coolant Pump
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- (NRC) Regulatory Guide
SBLOCA - Small Break Loss of Coolant Accident
- Short Term Instruction
TS
- Technical Specification
- Unre!.alved item
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