ML18058B832
| ML18058B832 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/24/1993 |
| From: | Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18058B833 | List: |
| References | |
| NUDOCS 9306010050 | |
| Download: ML18058B832 (15) | |
See also: IR 05000255/1993010
Text
Docket No.
50-255
Consumers Power Company
ATTN:
Mr. Gerald B. Slade
.
General Manager
Pal.isades Nuclear Plant
27780 Blue Star Memorial Highway
Covert, MI
49043-9530 -
Dear Mr. Slade: *
SUBJECT:
NRC INSPECTION REPORT NO. 50-255/93010(DRS)
This refers to the inspection .conducted 'by Messrs. J. Wa 1 ker and W. Pegg of *
this office on April 6 through May 4, 1993.
The inspettion included ~ revi~w
of authorized activities .for your Palisad~s Nuclear Plant. This inspection
focused on the implementation of procedural changes and training modifications
described by NRC Safety Evaluation Report (SER) dated February 28, 1986,
- regarding the single failure issue for main steam isolation valves (MSIVs) .
The operability and maintenance history of the *MsIVs were also examined.
At
the conclusion of the inspection, the findings were discussed with those
members of your staff identified in the entlosed report~
Areas examined during the inspection are identified in the report.
The
inspectors selectively examined procedures and representative records; made
observations, conducted interviews with your personnel, and conducted two
simulator runs of a dual steam generator blowdown concurrent with a failure of
. the intact steam generator's MSIV to close with operations personnel.
.
.
.-
No violations of NRC requirements were identified during the course of the -
inspection.
However, certain other activities, set forth in the enclosure to
this letter, appear to be a deviation.from commitments which you have made in
previous correspondence with the Commission.
A written response is required.
In addition~ certain actions taken in response to the 1986 SER did not fully * *
address the concerns expressed by the NRC.
These actions *should be reexamined
to erisur~ the intent of the SER has been met~ lhese are considered open *
items. A written response is requested within 60 days of the da~e of this
report to address the open items identified in this report.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of
this letter, the enclosures, and your responses to the letter will be placed
in the NRC Public Document Room.
The responses directed by this letter and the accompanying Notice are not
subject to the clearance procedures of the Office of Management and Budget as
required by the Paperwork Reduction Act of 1980, PL 96-511.
9306010050 930524
ADOCK 05000255
G
'PDR
Consumers Power Company
2
. May 24, 1993
We will gladly discuss any-questions you have concernjng this inspection.
Enclosures:
1.
Notice of Deviation
- . 2.
Inspection Report
No. 50-255/93010(DRS}
cc w/enclosures:
David P. Hoffman, Vice President
Nuclear Operations
OC/LFDCB .
Resident Inspector, Riii
James R. Padgett, Michigan Public
Service Commission
Michigan Department of *
Public Health
A. H. Hsia, LPM, NRR
SRI, Big ,Rock Point
W. Hodges, RI
. A. Gibson, RII
S. Collins, RIV
K. Perkins, RV
bee w/enclosures:
PUBLIC-IEOl
Sincerely,
original signed by
T. 0. Martin, Acting Director
Division of Reactor Safety
SEE PREVIOOS CXJNCURRENCE PAGE .
Riii
Pegg/jw/wp/cg
05/ /93
Riii
Walker
05/ /93
RI II
Burdick
05/ /93
Riii
Jorgensen
05/ /93 ~
Ring
Riii
VW'
Martin
05/'L~ /93
05/ 'l,~93
Consumers Power Company
2
. .
. -
.
We will gladly discuss any ~uestion~ you have concernJng this inspection.
Enclosures:
1.
Notice of Deviation
2.
Inspection Report
No. 50-255/93010(DRS)
cc w/enclosures:
David P. Hoffman, Vice President
Nuclear Operations
OC/LFOCB
Resident Inspector, Riii
James R .. Pa*dget t, Mi chi gan Public
Service Commission
Michigan Department of
Public Health
A. H. Hsia, LPM, NRR
SRI, Big Rock Point
bee:
PUBLIC - IEOl
R~J)~1
PeggtJ~/wp/cg
05/{)i/93
Rj].I
(.;';er
05/zi/93
Sincerely,
T. 0. Martin, Acting Director
- Division of React~r Safety
t.sen
Riii
.Ring
05/"--\\/93
05/ /93
Riii
Martin
05/ /93
NOTICE OF DEVIATION
Consumers Power Company
Palisades Nuclear Plant
Docket No.
50-255
License No. DPR-20
During* an NRC inspection conducted on April 6 through May 4, 1993, a deviation
of your written response to a safety evaluation report (SER) dated
February 28, 1986; regarding the single failure issue for main steam isolation
valves was identified.
In accordance with the "General Statement bf Policy
and Procedure for NRC Enforcement Action", 10 CFR Part 2, Appendix C, the
deviation is listed below:
In an April 28, 1986 letter to the NRC, the licensee committed, in part, to
perform certain actions for a.main steam line break coincident with failure of
the inlact steam generator's main steam isolation valve to close. These
specific actions include opening the power-operated relief valves, maximizing
safety injection flow (feed and bleed}, and maximizing auxiliary feed flow to
one steam generator if the decision is made not to rely on any in-containment
instrumentation.
In the basis docu~ent for emergency operating procedures (EOP) 9.0,
"Functional Recovery Procedure," and EOP 6.0, "Excessive Stearri Demand Event,"
the licensee made the conservative assumption that no in-containment
instrumentation would remain operable.
Contrary to the above, as of May 4, 1993, the emerg~ncy operating procedures
do not decisively instruct the operators to perform the above actions.
Instead, EOP 9.0 ambiguously directs the operators to .monitor in-containment
parameters, which may be inoperable or unreliable, to determine the proper
mitigating procedure.
Please provide to the U. S. Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington, D. C. 20555i with a copy to the Regional
Administr~tor, Region III, in writing within 30 days of the date of this
- Notice, the reason for the deviation, the corrective ~teps which have been
taken and the results achieved, the corrective ste~s which will be taken to
avoid further deviations, and the date when your corrective actibn will be
completed.
Where ~ood cause is shown, consideration will be given to
extending the response time.
Dated at Glen Ellyn, 'Illinois *
this 24th
day of May *, 1993
. 9306010055 930524
- PDR
ADOCK 05000255
Q
. </.o.JAn"~ . *
T. 0. Martln; Actrng Director
Division of Reactor Safety
U. S. NUCLEAR REGULATORY COMMISSION
REGION I I I .
- Report No. 50-255/93010(0RS)
Docket No. 50-255
Licensee:
Consumers Power Company
Palisades Nuclear Generating Plant
27780 Blue Star Memorial Highway
Covert, MI
49043-9530 *
Facility Name:. Palisades Nuclear Plant
Inspection At:
Covert, Michigan
License No. OPR-20
Inspection Conducted:
April 6 through May 4, 1993
~
Inspectors:. J. Walker
.
{-tn._
~~
Inspection Summary
~t-%'1/fJ .
Date
.
Inspection on April 6 through May 4. 1993 (Report No. 50-255/93010CDRSll
Areas Inspected:
Special inspection to assess the implementation of.
commitments made in response to the 1986 SER and tb*assess the bperability and
availability of the MSIVs.
.
Results: Within the areas inspected, one deviatiort was identified regarding a
conflict between the EOPs dealing with a main steam line break (MSLB)
coincident with a single failure of the intact steam generator's MSIV to close
and commitments made to the NRC (Section 2.4.6.1).
Open items were identified
re~arding differences between the training of operators and the
recommendations of the SER on the possibie loss of instrumentation inside of
containment during this event (Section 2.4.6.2); a commitment to include as-
found stroke time testing of the MSIVs in the operational procedures (Section
2.4.4); operator training on loss of instrumentation and procedural changes
related to this ~vent were not develriped; graphs to account for errors in
pre~surizer and SG narrow range le~el were referred to in footnotes in the
safety status, but no guidance is given as to when these graphs are required
(Section 2.4.7.1); and discrepancies between the containment. pressure and.
temperature values observed during simulator runs of the MSLB event and the
. expected values (Section 2.4.7.3).
No problems with the maintenance or
operability of the MSIVs were noted.
"9306010060 930524
ADOCK 05000255
G
-
II
DETAILS
1.0
Exit Meeting Attendees
2.0
2.1
Consumers Power Company
T. Palmisano, Operations Manager
R. Orosz, Manager, Nuclear Engineering and Construction
J. Hanson, Operations Superintendent
R. Heimsath, Training Administrator
J~ Kuemin, Licensing Administrator
. R~ Kasper, Maintenance Manager
R. Gerling, Reactor and Safety Analysis Manager
D. Malone, Operations Staff Support Supervisor
- F. Yanik, Risk Assessment Supervisor
B. Kubacki, System Engineering Section Head
E. Letke, Main Steam System Engineer
U. S. Nuclear Regulatory Commission {NRC) *
T. Burdick, Chief, Operator Licensing Secti6n 2
J. Heller, Senior Resident Inspector
The NRC inspectors also contacted and interviewed other licensee
personnel during the inspection.
Background Information
Purpose of the Inspection
The purpose of the inspection was to assess implementation of commitments made
i~ response to the 1986 SER.and to assess the operability and availability of
the MSIVs through an examination of the MSIV maintenance and operational
history.
2.2
History
In the early 198d's, it was dis~overed that the design of the main steam
system did not preclµde the possibility that a MSLB inside containment
concurrent with a failure of the intact steam generator's (SG) MSIV could
occur.
Th~ steam system at.Palisades relies on.reverse mounted check valves
as MSIVs.
There are no non~return check valves downstream of the MSIVs to
prevent blow-down of both steam generators in the event of a steam line break
inside containment concurrent with a passive failure of the opposite steam
gener.ator's MSIV to close.
In 1983, the licensee committed to make modifications to either replace the
MSIVs with motor-operated valves, install non-return check valves downstream
of the MSIVs to preclude backflow from one steam generator into the other on a
MSLB, or qualify the instrumentation and equipment inside of containment for
the bounding .conditions of the dual steam generator blowdown.
2
In 1985, the licensee submitted a PRA to the NRC to show that the proposed
modifications were not cost effective considering the probability of the event
occurring and the cost of the improvements relative to the decrease in overall
risk for each of the propose.d modifications.
On February 28, 1986, the NRC issued an SER on this issue concurring with the
licensee's evaluation provided alternative measures were implemented.
These
alternative measures, the licensee's response to theni and the NRC's evaluation
of the response are provided in Section 2.4.6 of this report .
. 2.3
Valve Design
The Palisades MSIVs are 30-inch Atwood-Morrill swing di~c check valves that
are reverse mounted and utilize an external air cylind~r and piston connected
to the valve disc shaft to hold the valves open.
To close the MSlVs, a
solenoid is actuated which vents the air from under the piston. This allows a
compressed spring above the piston t~ expand ~nd force the piston down,
assisting th~ valve in closing.
. 2 .4
2.4.l
Inspection Findings
Evaluation of Industry Experience
On September 29, 1991, while.in the process of shutting down, the MSIVs at the
Point Beach Nuclear Plant, Unit 1, failed to close due to corrosion problems
in the*pneumatic oper~tors. On November 15, 1991, the lic~nsee began a review
of .this industry ~xperience and determined that the Palisades plant had eight
Atwood~Morrill check valves of a similar de~ign, including the MSIVs.
Prior
to evaluation of the MSIV failure's* applicability to Palisades, the licensee's
document tracking the industry experience was lost due a computer .entry error
and a failure of the responsible review group to return the evaluation to the
originator.
The review was started again on March 11, 1993, after NRC
operations inspectors made inquiries about the MSIVs and the.event at Point
Beach.
Consequently, the licensee did not complete an evalu~tion of the
applicability of the Point Beach experience to Palisades in a time frame
commensurate with the safety significance.
The control exhibited over this
review is considered a weakness. *
- 2.4.2
MSIV Operational History
The inspectors reviewed licensee event reports, abnormal occurrence reports,
and interviewed operations personnel to determine if the MSIVs had ever failed
to close on demand.
The inspectors determined:that the MSIVs had failed to
c]os~ three times in the period between 1972 and 1973.
There have been no
failures to close since 1973.
The actions taken in response to the initial
failures were successful in.preventing recurrence.
2.4.3
MSIV Maintenance History
The inspectors reviewed work orders from the past ten years, with emphasis on
maintenance activities over the past three years.
In Work Order 24104695, the
. pneumatic operator of CV-0510 was disassembled and inspected per Procedure
3
MSS-M-19, "Disassembly, Inspection and Reassemb1y of Main Ste~m Isolation
Valves CV-0501 and CV-0510", Revision 5.
The inspection of the actuator
.
- internals revealed some spots of corrosion, scratches and other normal wear.
Th~ degree of wear and degradation discovered would not have affected the
ability of the actuator to perform it's function.
The.licensee will
disassemble and inspect the pneumatic actuator for CV-0501 during the next
. cold shutdown.
Although prbvlsions are made in this procedure to perform
disassembly and inspection of the pneumatic actuator (step 5.3.9), step 3~8.4
allows the system engineer to bypass this step of the procedure.
This step in
the procedure had been bypassed since the installation of the actuators in
1979.
The length of* time between the. installation of the pneumatic actuators
in 1979 and the first disassembly and inspection of the actuators in 1992 is
excessive considering the importance of the actuators and is a weakness.
Th~ inspectors were concerned about the use of Q Sealant, Furmanite, or other
temporary means to repair stuffing box and packing leak.s without the benefit
of as~found testing to determine if such activities unduly affected the
performance of the MSIVs.
As-found testing would provide necessary feedback
- On the impact of such repairs on the shaft friction and the ability of the
actuator to overcome the increased friction~
The inspectors found the maintenance procedures to be clear and complete .. The
inspectors interviewed maintenance personnel and determined that the
maintenance personnel were qualified and cognizant of the maintenance
requirements of the MSIVs.
No other problems were noted.
2.4.4
MSIV Testing
The inspectors reviewed the technical specification (TS) surveillance history
of the MSIVs for the past five years.
The MSIVs closed within the TS
allowable times and conditions.
Procedure Rl-17, "Main Steam Isolation Valve
Circuits Test and Valve Closure limingri, Revision 12J documented the required
TS closure time testing. The required testing, however, did nbt accurately
reflect the ability ~f the MSIVs to perform their safety function during the
previous operational cycle, as the MSIVs were te~ted on power ascension after
they had already been exercisedJ repaired or conditioned.
Prior to this inspection,-the licensee had documented closure of the MSIVs on
a shut down check list, but no measurement of the closure time was taken.
The
licensee has since committed to include as-found stroke time testing of the
MSIVs in the operational procedures.
The results of this as-found testing
will indicate if there was any degradation in the ability of the.MSIVs t~
perform their function over the previous operational cycle.
The commitment to
. incltide as-fo~nd stroke time testing of the MSIVs in the operation procedures
will be tracked as an open it~m (255/93010-0l(DRS)).
2.4.5
Dual SG Slowdown
The following is a brief synopsis of what would happen in a dual SG blowdown,
based on the PRA and simulator runs of this scenario on the Palisades
certified plant specific simulator.
The use of the designators "A" and "B"
are arbitrary and are provided for clarity.
4
2.4.6.
2.4.6 .. 1
A guillotine break occurs on the B Main ~team line, upstream of
the MSIV.
The MSIVs receive a low steam generator pressure signal to close.
The MSIV for the A steam generator fails ~o close and remains in.
the full open position.
.
.
Backfl ow from the A steam generator pushes the* MSIV for the B
stea~ generator ripen and allows the A steam gen~rator to blow down
through the b_reak.
Both steam generato~s blow down into containment.
The primary coolant system (PCS) cools down to approximately 350°F
and depressurizes to approximately 1000 psia. All primary coolant
pumps (PCPs) are stopped.
Upon dry out of the ~Most Affected" _steam generator the PCS begins
to reheat and repress~rize. The saturation t~mperature in the
"Least Affected" steam generator is initially above PCS
temperature preventing it from acting as a heat sink and
inhibiting natural circulation.
Voids are formed in*the upper head region of the reactor vessel,
again inhibiting natural circulation flow.
Licensee Response to SER
Operator Actions
The NRC concern, documented in the SER issued in 19S6, was as foll~ws:
"The emergency procedures dealing with secondary line breaks, EOP-6 and
EOP-7, and the procedure for normal reactor trip, EOP-1, do not provide**
definitive guidance regarding maintenance of a heat sink (use of steam
generator level or other secondary parameter for feedback) even if thi
operator perceives that both steam generators may be faulted. In fact,
in EOP-1 it is noted that if dry out occurs, the affected steam
generate~ is to be considered inoperable. Additionally, the absence in
. the procedures of ~ny rec6gnition of overcooling expected to occur in
these events enhances the potential for inappropriate spontaneous
operator action. Procedures based on systems analysis should be
developed to enable the operator to cope with these events and to .
achieve a controlled cooldown: T~e staff considers that -analyses are
needed to identify a control strategy that.considers the decay heat
generation available to the operator and how the response changes over
the course of the cooldown." *
5
The licensee's response to this concern, documented in a letter to' the NRC
dated April
28~ 1986, was as follows:
~All new*EOPs provide definiti~e guidance ~egarding maintenance of the
heat sink. This includes monitoring of specific parameters including
wide range steam generator levels, core temperature indications, steam
generator pressures and st~am generator feed flow.
All
ne~ EOP~ have
safety function status checks whichinclude the safety function of core
heat removal and PCS heat removal."
"In all new EOPs, emphasis is placed on continuous feed to at least one
steam generator, even if both SGs are "dried out", until shutdown
cooling or long term-po~t PCS break-core cooling is sufficient for decay
heat removal."
"It should also be emphasized that in all new EOPs, confirmation of
adequate core heat removal requires the confirmatory indication of at
1 east in-core temperature.
In the Funct i ona 1 Recovery Procedure, if in-
coritai nment indications are consi~ered lost, the continuing actions for
PCS Heat Removal *and Core Heat Removal include maximizing feed flow to
available steam generators."
"The new EOPs are structured such that, if in-containment vital
instrumentation is considered lost then the Functional Recovery
- Procedure--is implemented.
If* a decision is ~ade to not rely on any in-
containment instrumentation, the result is that (1) the PORVs are opened
and all available safety injection (SI) pumps started and SI valves
. *
open; (2) auxiliary feed flow is maximized and directed to at least one
S/G;. (3) containment *air coolers are placed in emergency configuration
- and containment spray manually maximized; and (4) all avai-lable servic~ -
water and component cooling water pumps operated."
In developing a strategy to recover from this eventi the licensee made the
a~sumption that all in-containment instrumentation would be lost (Palisades
Plant Emergency Operatfng Basis Document, EOP 6.0 page 38 of 41, EOP 9.0 page-
25 of 133 and PRA,* Appendix 6, page 2). This would require that the strategy
~xplained above be performed.
Initially,. the operators would complete EOP 1.0, "Standard Post T~ip Actions."
Up6n completion of these actions, the operator would either initially ente~
EOP 6.0, "Excessive Steam Demand Event," or go directly tb.EOP 9.0,
"Functional Recovery Procedure," if*the operator determines that more than one
event is in progress.
Upon ente~ing EOP 9.0, the operator fs directed to Succe~s P~th HR-3, "S/G
heat sink and safety injection pump operation," based on actual S/G level
being 1 es*s than - 84% of wide range steam generator 1 eve ls. _ The fo 11 owing
actions are required by.HR-3:
1)
Wheri PCS pressure falls below 1300 psia, trip all PCPs. [PCS.
pressure will decrease to below 1300 psia due to the rapid
cooldown of the PCS caused by the steam break .. The functional
6
2)
3)
4)
5)
6)
recovery procedure, not assuming any specific single event is in
progress, has the PCPs tripped in the event a LOCA could be
occurring];
Opetate all safety injection and chargin~ pumps;
!so.late the MOST affected steam generator, [this requires the
closing of both MSIVs which cannot be accomplished];
Verify feedwater criteria by either steam generator level
(steam generator level is one of the instruments assumed lost by
this event], or feedwater flow, in conjunction with steam
generator levels [not available];
Check for acceptance criteria of HR-3 by SG level [not available],
core exit thermocouples or loop cold leg temperature, [not
available], and adequate saf~ty injection flow~ and
If HR-3 acceptance criteria is not met, then go to HR-4, "Once-
Through-Cool ing."
[By using instrumerit~ inside containment ~hich
are not reliable, the operator may not be implementing once-
through-cooling ~t this point in the procedure.]
Decisions made in u~ing the EOPs as written rely upon in~trumentation that has
alteady been assumed to be inoperable and/or unreliable.
The PRA agrees that
in-contain~ent instrumentation will be lost along with in-cont~inment
electrical components.
"Violation of EEQ is assumed to lead to failure of
instrumentation required by the operator to monitor systems and equipment
within containment important to maintaining adequate core cooling."
(PRA
Appendix 6, Page 2)
The licensee continues to take credit for the. in-containment instrumentation
and electrical components when using EOP 6.0, "Excessive Steam Demand Event,"
and EOP 9.0, "Functional Recovery Procedure," for recovery.
The ltcens~e maintains the PORVs closed with their block v~lves isolated.
Ther~ i~ no guarantee that the block ~alves would be able tQ be opened in the
harsh environment inside contairtment.
"Currently to comply with 10 CFR 50,
Appendix R, PORV breakers52-196 and 52-224 are.normally left open.
This is
to:prevent the po~sibility of "hot shorts" simultaneously in control circuits
for each PORV and its associated block valv~. resulting in a LOCA."
(Supplement 1 to NUREG-0737-Response to Draft Safety Evaluation - Procedures
Generation Package, Attachment 1, page 3).
The licensee has failed to meet their commitment with regard to the following:
(1) go to once-through-cooling; (2) ensure.maximum feed flow to at least one
steam generator;. (3) maximize containment. spray flow and place containment* .
coolers in emergency alignment; and (4) maximize service water and component
- cooling water flow.
Due to the facility relying on inoperable or unreliable
instrumentation, there is no guarantee that the steps committed to would be
performed.
With the potential loss of instrumentation and other components
upon diagnosis of the dual steam generator blowdown with failed open MSIV and
7
upon exiting_ EOP 1.0, the operator should immediately go to once through
cooling to ensure the core remains cooled, then make atte~pts to regain SGs as
a viable heat sink. This is considered a deviation from the specified
commitment
(255/93010-02(DRS~ 2.4.6.2 Loss of Instrumentati~n The NRC concern, documented in the SER issued in 1986, was as follows: "Because some instrumentation inside contai~ment may be affected by the two.steam generator blowdown environment, operator training/procedures to cope with possible loss of information or misinformation is needed."* The licensee~s response to this toncern, documented in a letter to the NRC dated April 28, 1986, was as follows: "Caution notes will be added to the beginning of operator actions and
- Safety Function Status checks of the optimal recovery procedures for
LOCA-and Excess Steam Demand Event (ESDE-the events in which degraded containment atmospherit conditions are expected}. These notes will state that due to the potential adver~e effects of high containment temperature/pressure on in-containment instrumentation the operator should not rely on any single -instrument, but rather observe confirmatory 1ndications; *and significance should be placed heavily on trending and less significan~e on s~ecific instrument readings * "In light bf these notes, the existing guidahce in the.LOCA and* ESDE draft procedures for maintaining SG feed flow would prevent deliberately stopping auxiliary feedwater flow. *Additionally, the SI pump throttling criteria; which require a high operator confidence level in primary temperature and pressure indications, would correctly prevent any operator action limiting SI flow.
- "The acceptance ~riteria for containment atmospheric conditions in the
LOCA and ESDE procedures will be chang~d to have an upper limit of design pressure and temperature. Therefore, these conditions in safety function status check should "flag" operators into using the *Function~l R~covery Procedure":
The licensee, in stating that a caution note would be added to the EOPs, took credit for an existing caution. "During degr~ded Containment conditions, the * operator should not rely on any single instrument indication due to large instrument errors; Alternate/additional instrumentation should be used to confirm trending of PCS conditions." This caution does not give the operator any guidance as to what instrumentation is lost or what alternate instrumentation is available. In addition, the ba~is documents for EOP 6.0 and EOP 9.0 state that all containment instrumentation is assumed lost for the MSLB with single MSIV failure scenario (EOP .6.0 Basis, Revision 4, page 38 of 41 and EOP 9.0 Basis, Revision 5, page 25 of 133). 8
,: The licensee met their commitment in this case, but failed to fully address the concerns expressed by the NRC. In particular, adequate procedure actions. to cope with the loss of information or misinformation caused by the loss of instrumentation were not established. The inspectors reviewed the training records related to this event and found no records of training on loss of instrumentatfon. Some guidance ts given in EOP 6.0 for narrow range steam generator level and pressuriier level, but this would only be applicable for a single uncomplicated steam line break (no other events in progress concurrent with the steam line break). Operations personnel stated that they did ~ot know whit instrumentation would be affected, and that they would have to wait for the engineering staff to evaluate containment conditions and determine which instrumentation would be available for use. This evaluation would take place during the event. The licensee failed to fully address the concerns raised by the NRC in. . response to item 2. Specifically, adequate changes to *the procedures and operator training on loss of instrumentation related. to this event were not .developed. This is considered an open item (255/93010-03(DRS)). 2.4.7 2.4.7.1 'Additional Areas Examined Emergency Ope rat i nq Proce*dures The cauti6n prior to step 12 in EOP 9.0, Success Path 9.0, states ".If maintaining an* SG heat sink is immediately needed to protect the Public Health/Safety and both SGs are required to be isolated, then the Shift Supervisor may direct dep~rting from the SG isolation steps (reference 10CFR50~54X)~" The basis for this caution states "If both SGs are affected, isolating both SGs will result in a violation of HR-3 safety function acceptance criteria. Only HR-4, "Once-Through-Cooling," would be available". This caution should direct the operators to the proper pr6cedure, HR-4 .. The caution, stating that the Shift Supervisor may deviate from the procedure via.* 10 CFR 50.54x, -provides no useful information. In this case, actions. that can provide adequate protection are not specified even though they are available. In addition, 10 CFR 50.54x is invoked in an emergency when r~asonable actions that depart from a li~ense condit~on or a technical specification are needed to protect the public health and safety and no actions consistent with license conditions and technical specifications that can provide adequate or - ~quivalent protection are immediately apparent. In this case, .appropriate actions* can be taken through HR-4. EOP 6.0, Attachment 2, contains graphs to account for errors in pressurizer and SG narrow range level: This attachment in not referred to in the body of the ~rocedure. It is referred to in footnotes in the Safety Status~ but no guidance is given as to when these graphs are required. This was brought to the licensee's attention with no immediate response. This is considered an open item (255/93010-04(DRS)). .9
2.4.7.2 Training The inspectors reviewed the training for this event and determined that the only training provided was on EOP 6.0. This training primarily consisted of being able to identify the "Most* Affected" and "Least. Affected" steam .. generators. It did not cover the backgiound to this event or the tommitments made to the NRC as to what actions were necessary to recover from this event. No training was conducted on th~ potential loss of instrumentation or actions needed to be taken in response to the loss of instrumentation. No training was conducted on EOP 9.0, which is the facility identified primary procedure for recovering from this event.
. . During the 1989 Emergency Ope~ating Procedure Team Inspection, it was identified that, "there were no lesson plans for individual success path . procedures of EOP 9.0, either in the simulator or the classroom phases of instruction." In the-current training material, the lesson plan for EOP 9.0
- covered the generic steps for EOP 9.0 and addressed the basis for each suctess
path in an overview format. Individual training for each success path was not performed. In discussions with the training staff, the inspec~ors noted that d~ring the previous training cycle, a scenario had been run to cover this event. The ' following items were noted in a review of the Simulator Exercise Guide. 2.4.7.3 Plant trip was initiated by the inadvertent closure of the MSIVs with failure of one MSIV to fully close. The training staff inserted leaks on both steam ge~erators to * simulate a MSLB in containment concurrent with failure of one MSIV. Thisw.as done in conjunction with operator over rides on the MSIV indications to give the appearance of. the failed open MSIV. The dynamics of this event differed from those obtained by running a dual steam generator blowdown with one MSIV*failed open. . . . The criteria for ending this scenario was, at a minimum, for the operators to enter EOP 9.0 and determine which success.path to use. The scenario could be allowed to continue until the most affected SG is isolated or until the crew had entered HR-4, "Once-Through-Cooling". No training was conducted on the potential loss of-ihstrument~tion inside containment. Observations of Simulator Exercise The following parameters did not appear to be properly modeled on the Palisades Plant specific simulator. Containment Pressure ~ During the first scenario with all ECCS equipme_nt available, the containment pressure peaked at 12 psig. During the ~econd scenario with a loss of off site power and only one train of
- ECCS equipment available, maximum containment pressure was_22 psig.
10
Previous analyses iridicated that these pressures are lower than what would be present during an actual event. Containment Temperature - The maximum containment temperature identified during either scenario was 150°F. Previou~ analyses indicated that this temperature is lower than what would be pre~ent during an actual event.
As a result of the containment temperature and pressure discrepancies, it is believed that the training on this event had a negative effect. The operators were not presented with realistic*values to enable them to make the required decisions. This tends to lead to a false sense of security and reliance on procedures that do not appear adequate to recover from this event. This ~s considered an open item (255/93010~05(DRS)). 3.0 Open Items Open items are matters ~hich have been discussed with the licensee, ~hich will be reviewed further by the inspectors, and which involve some action on the part of the NRC or licensee or both. Open items disclosed during the inspection are discussed in Sections 2.4.4, 2.4.6.2, 2~4.7.1, and 2.4.7.3 of - this report. 4.0 Exit Meeting The inspectors met with the licensee staff (denoted in Section 1) at the site on May 4, 1993, for an exit meeting to summarize the purpose, scope, and findings of the inspection. A verbal summary of the inspection findings was provided to the licensee at that time. The inspectors discussed the likely informational content of the *inspection report with regard to documents or processes reviewed by the inspectors during the inspectiori. * The licensee did* not identify any such documents or processes as proprietary. 11 }}