Preliminary Clarification of TMI Action Plan Requirements - Addendum to 9/5/80 Letter (Generic Letter 80-81)
GL80081
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555
September 19 1980
TO: All Licensees of Operating Reactor Plants and Applicants for
Operating Licenses and Holders of Construction Permits
SUBJECT: ADDENDUM TO THE CLARIFICATION LETTER FOR TMI ACTION PLAN
REQUIREMENTS
By letter dated September 5, 1980, we transmitted a preliminary
clarification of the TMI Action Plan requirements. Attached is a set of
errata sheets which amend the referenced letter (viz., missing pages,
scheduling, Tech Spec consideration, etc.). Also included is a corrected
table of the Implementation schedule.
It is our intention to develop and issue model technical specifications
after issuance of the final requirements package.
Sincerely,
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
Enclosure:
As stated
cc w/enclosure
.
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(11.F.2)
d. An evaluation, including proposed actions, on the conformance of
the inadequate core cooling instrumentation system to Regulatory
Guide 1.97, Rev. 2. Any deviations should be justified.
e. A description of the computer functions associated with ICC
monitoring and functional specifications for relevant software in
the process computer and other pertinent calculators. The
reliability of nonredundant computers used in the system should be
addressed.
f. A current schedule, including contingencies, for installation,
testing and calibration, and implementation of any proposed new
instrumentation or information displays.
g. Guidelines for use of the additional instrumentation, and analyses
used to develop these procedures.
h. A summary of key operator action instructions in the current
emergency procedures for inadequate core cooling and a description
of how these procedures will be modified when the final monitoring
system is implemented.
i. A description and schedule commitment for any additional
submittals which are needed to support the acceptability of the
proposed final instrumentation system and emergency procedures for
inadequate core cooling.
TECHNICAL SPECIFICATION CHANGES REQUIRED
Yes.
REFERENCES
1. NUREG-0578 (Recommendation 2.1.3.b).
2. H. Denton (NRC) letter to All Operating Nuclear Power Plants on
"Discussion of Lessons Learned Short Term Requirements," dated October
30, 1979.
.
EMERGENCY POWER FOR PRESSURIZER EQUIPMENT
(II.G.1)
POSITION
Consistent with satisfying the requirements of General Design Criteria 10,
14, 15, 17 and 20 of Appendix A to 10 CFR Part 50 for the event of
loss-of-offsite power, the following positions shall be implemented:
1. Motive and control components of the power-operated relief valves
(PORVs) shall be capable of being supplied from either the offsite
power source or the emergency power source when the offsite power is
not available.
2. Motive and control components associated with the PORV block valves
shall be capable of being supplied from either the offsite power source
or the emergency power source when the offsite power is not available.
3. Motive and control power connections to the emergency buses for the
PORVs and their associated block valves shall be through devices that
have been qualified in accordance with safety-grade requirements.
4. The pressurizer level indication instrument channels shall be powered
from the vital instrument buses. The buses shall have the capability of
being supplied from either the offsite power source or the emergency
power source when offsite power is not available.
CLARIFICATION
1. While the prevalent consideration from TMI Lessons Learned is being
able to close the PORV/block valves, the design should retain, to the
extent practical, the capability to open these valves.
2. The motive and control power for the block valve should be supplied
from an emergency power bus different from that which supplies the
PORV.
3. Any changeover of the PORV and block valve motive and control power
from the normal offsite power to the emergency onsite power is to be
accomplished manually in the control room.
4. Far those designs where instrument air is needed for operation, the
electrical power supply requirements should be capable of being
manually connected to the emergency power sources.
APPLICABILITY
All PWR Operating License Applicants
.
CONTROL OF AFW INDEPENDENT OF ICS
(II.K.2.2)
POSITION
For B & W designed reactors, provide procedures and training to initiate and
control auxiliary feedwater independent of the integrated control system
(ISC).
CLARIFICATION
None required
APPLICABILITY
All Operating License Applicants with B & W designed reactors
IMPLEMENTATION
Prior to issuance of a full power license
DOCUMENTATION REQUIRED
Applicants shall provide sufficient documentation at leat four months prior
to the issuance of a full power license to support a reasonable assurance
finding by the NRC that the position specified above has been met.
TECHNICAL SPECIFICATIONS REQUIRED
No.
REFERENCES
NUREG-0660, (Section II.K.2, Table C.2, Item 2)
NUREG-0694, (Part II)
.
AUXILIARY FEEDWATER SYSTEM UPGRADING
(II.K.2.8)
POSITION
All operating Babcock and Wilcox plants were ordered to be shut down shortly
after the TMI-2 accident. The Orders included both short-term and long-term
actions. The NRR Bulletins and Orders Task Force reviewed the licensees
compliance with the short-term actions of the Orders and issued safety
evaluation reports which served as the basis for plant restart. Additional
items were identified in the review of the long-term actions which requires
further work by the licensees.
CLARIFICATION
The licensees were required to comply with the Commission Orders regarding
certain short-term and long-term AFWS modifications. The staff evaluated the
short-term actions, and safety evaluations were prepared prior to the plants
being allowed to return to operation. The staff evaluation of the additional
(long-term) items will be performed in conjunction with Item II.E.1.1,
(Auxiliary Feedwater System Evaluation).
APPLICABILITY
All B&W Operating Reactors.
IMPLEMENTATION
No separate implementation is required for this item. All AFW system upgrade
modifications for B&W plants are being reviewed as part of Item II.E.1.1.
TYPE OF REVIEW
See Item II.E.1.1.
DOCUMENTATION REQUIRED
See Item II.E.1.1.
TECHNICAL SPECIFICATION CHANGES REQUIRED
As required.
REFERENCES
NUREG-0660, (Sections II.E.1.1. and II.K.2.)
NUREG-0645, Volume 1, (Section 2.4.6)
.
FAILURE MODE EFFECTS ANALYSIS ON ICS
(II.K.2.9)
POSITION
B&W licensees submit a failure mode and effects analysis (FMEA) of the
integrated control system (ICS).
CLARIFICATION
A generic FMEA of the ICS (BAW-1564) was submitted on August 17, 1979 by the
operating plant licensees. This report was reviewed by the staff and ORNL.
Requests for additional information, regarding the recommendations contained
in the report, were sent to the licensees on November 7, 1979. The responses
to the November 7 1979 letter have been received and are under review.
APPLICABILITY
All B&W Operating Reactors and Operating License Applicants.
IMPLEMENTATION
Operating Reactors
Open - Staff recommendations pending completion of staff review.
Operation License Applicants
Prior to issuance of a full-power license.
TYPE OF REVIEW
Postimplementation review.
DOCUMENTATION REQUIRED
Operating Reactors
To be determined following staff review.
Operating License Applicants
B&W applicants should provide the following:
1. Identify whether the previous generic submittal (BAW-1564) is
applicable to your plant, and
2. Specify what actions have been taken at your facility to comply with
the recommendations listed in BAW-1564.
.
SAFETY-GRADE ANTICIPATORY REACTOR TRIP
(II.K.2.10)
POSITION
Upgrade the currently installed control-grade, anticipatory reactor trip
(ART) on loss-of-feedwater and turbine trip to safety-grade.
CLARIFICATION
Operation Reactors
1. IE Bulletin 79-058, Item 5, issued on April 21, 1979, directed B&W
licensees to provide a design and schedule for implementation of a
safety-grade reactor trip upon:
a. loss of feedwater;
b. turbine trip; and
c. significant reduction in steam generator level.
2. In accordance with IE Bulletin 79-058, the B&W licensees submitted a
conceptual design for a safety-grade, anticipatory reactor trip which
would be initiated upon turbine trip and loss of feedwater only.
Included in the licensees' responses was a generic evaluation prepared
by B&W which proposed that the anticipatory reactor trip on low steam
generator level was not necessary.
3. Staff review of these submittals resulted in a preliminary design
approval for the safety-grade. anticipatory reactor trip being issued
to the B&W licensees on December 20, 1979. However, the approval
letters also specified the additional information which would be
required to be submitted prior to final staff approval of the design.
4. The staff will complete its review of the generic evaluation by B&W
which indicates that the proposed anticipatory trip on low steam
generator level is unnecessary. Further clarification will be provided
on this matter, if required, following completion of the staff review.
Operating License Applicants
Compliance with TMI Action Plan, Item II.K.1.21, satisfies this requirement.
APPLICABILITY
All B&W Operating Reactors and Operating License Applicants.
.
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TECHNICAL SPECIFICATION CHANGES REQUIRED
Yes.
REFERENCES
Commission Orders on B&W Plants
IE Bulletin 79-05B, Item 5
Letter from R. W. Reid (NRC) to B&W Licensees, data December 20, 1979
Subject: Preliminary design approval for safety-grade anticipatory reactor
trip and request for additional information
NUREG-0645, (Volume 1, Section 2.4.6)
NUREG-0694, (Part 2)
.
CONTINUED OPERATOR TRAINING AND DRILLING
(II.K.2.11)
POSITION
Continue operator training and drilling to assure a high state of
preparedness.
CLARIFICATION
In a letter from D. F. Ross, Jr. (NRC) to All B&W Operating Plants, dated
August 21, 1979, each B&W licensee was requested to document the steps they
had taken to insure that continued operator training and drilling
incorporated the necessary lessons learned from the accident at TMI-2. This
response was required to assure compliance with the long-term training
requirements of the Commission Orders.
Responses to this request were received from the licensees and reviewed by
the NRC staff. Based on that review, the staff concluded that the training
programs had been sufficiently modified to incorporate the necessary lessons
learned from TMI such that this portion of the Commission Orders was
satisfied. A complete evaluation of this item is discussed in Section 2.4.6
of NUREG-0645, Volume 1.
Additional requirements, beyond the intent of the Commission Orders, are
being implemented through the following items of the Action Plan: I.A.2.2,
I.A.2.5, I.A.3.1, and I.G.1.
APPLICABILITY
All B&W Operating Reactors
IMPLEMENTATION DATE
COMPLETED
TYPE OF REVIEW
Postimplementation review.
DOCUMENTATION REQUIRED
No additional documentation required.
TECHNICAL SPECIFICATIONS REQUIRED
No.
.
THERMAL MECHANICAL REPORT-EFFECT OF HPI
ON VESSEL INTEGRITY FOR SMALL BREAK LOCA
WITH NO AFW
(II.K.2.13)
POSITION
Perform a detailed analysis of the thermal-mechanical conditions in the
reactor vessel during recovery from small breaks with an extended loss of
all feedwater.
CLARIFICATION
The position deals with the potential for thermal shock of B&W reactor
vessels resulting from cold safety infection flow. One aspect that bears
heavily on the effects of safety injection flow is the mixing of safety
injection water with reactor coolant in the reactor vessel. B&W has
committed to provide a report by the end of July which will discuss the
mixing question and the basis for a conservative analysis of the potential
for thermal shock to the reactor vessel.
APPLICABILITY
All B&W Operating Reactors and Operating License Applicants.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of any modifications will
be subject to the results of NRC staff review of the report.
TYPE OF REVIEW
Post implementation Review
DOCUMENTATION REQUIRED
Licensees shall submit results of evaluation by January 1, 1981. Applicants
shall submit results of evaluation at least four months prior to the
issuance of a full power license.
TECHNICAL SPECIFICATION CHANGES REQUIRED
To be determine following staff review.
REFERENCE
NUREG-0645, (Volume 1 Section 2.4.5)
Letter from D. F. Ross Jr. (NRC) to all B&W Operating Plants, dated August
21, 1979.
.
EFFECTS OF SLUG FLOW ON STEM GENERATOR TUBES
(II.K.2.15)
POSITION
While the staff believed that the potential for slug flow was not great in
B&W plants, because of the venting path provided by the internal vent
valves, the staff required a confirmatory evaluation of the effects of slug
flow on steam generator tubes be performed by the licensees to assure that
the tubes could withstand any mechanical loading which could result from
slug flow.
CLARIFICATION
The request for this information was originally sent to the B&W licensees in
a letter from R. W. Reid (NRC) to All B&W Operating Plants dated November
21, 1979.
The results of this analysis has been submitted by the licensees and is
presently undergoing NRC staff review.
APPLICABILITY
All B&W Operating Reactors and Operating License Applicants.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of any modifications will
be subject to the results of NRC staff review of the evaluation.
TYPE OF REVIEW
Postimplementation.
DOCUMENTATION REQUIRED
No additional documentation is required at this time from Licensees.
Applicants must supply the requested information at least four months prior
to issuance of a full power license.
TECHNICAL SPECIFICATION CHANGES REQUIRED
No.
REFERENCES
Letter from R. W. Reid (NRC) to All B&W Operating Plants, dated November 21,
1979.
NUREG-0565, (Recommendation 2.6.2.1)
NUREG-0645, (Volume 1, Section 2.4.6)
NUREG-0694, (Part 2)
.
REACTOR COOLANT PUMP SEAL DAMAGE
(II.K.2.16)
POSITION
Evaluate the impact of reactor coolant pump seal damage and leakage due to
loss of seal cooling upon loss of offsite power. If damage cannot be
precluded, licensees should provide an analysis of the limiting small-break
LOCA with subsequent RCP seal damage.
CLARIFICATION
The request for this information was originally sent to the B&W licensees in
a letter from R. W. Reid (NRC) to All B&W Operating Plants dated November
21, 1979.
The results of these evaluations have been submitted by the licensees and
are presently undergoing NRC staff review.
APPLICABILITY
All B&W Operating Reactors and Operating License Applicants.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of any modifications will
be subject to the results of NRC staff review of the evaluations.
TYPE OF REVIEW
Postimplementation.
DOCUMENTATION REQUIRED
No additional documentation is required at this time from Licensees.
Applicants shall submit the requested information at least four months prior
to the issuance of a full power license.
TECHNICAL SPECIFICATION CHANGES REQUIRED
No.
REFERENCES
Letter from R. W. Reid (NRC) to All B&W Operating Plants, dated November 21,
1979.
NUREG-0565, (Recommendation 2.6.2.f)
NUREG-0645, (Volume 1, Section 2.4.6)
NUREG-0694 (Part 2)
.
POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS
(II.K.2.17)
POSITION
Analyze the potential for voiding in the reactor coolant system during
anticipated transients.
CLARIFICATION
The background for this concern and a request for this analysis was
originally sent to the B&W licensees in a letter from R. W. Reid (NRC) to
All B&W Operating Plants, dated January 9, 1980.
The results of this evaluation has been submitted by the B&W licensees and
is presently undergoing staff review.
APPLICABILITY
All B&W Operating Reactors.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of an modifications will
be subject to the results of NRC staff review of the licensees' evaluation.
TYPE OF REVIEW
Postimplementation Review.
DOCUMENTATION REQUIRED
No additional documentation is required at this time.
TECHNICAL SPECIFICATION CHANGES REQUIRED
No.
REFERENCES
Letter from R. W. Reid (NRC) to All B&W Operating Plants, dated January 9,
1980.
NUREG-0660, Section II.K.2 Item C.17.
.
SEQUENTIAL AFW FLOW ANALYSIS
(II.K.2.19)
POSITION
Provide a benchmark analysis of sequential auxiliary feedwater flow to the
steam generators following a loss of main feedwater.
CLARIFICATION
This requirement was originally sent to the B&W licensees in a letter from
D. F. Ross Jr. (NRC) to All B&W Operating Plants, dated August 21, 1979.
The results of this analysis has been submitted by the B&W licensees and is
presently undergoing staff review.
APPLICABILITY
All B&W Operating Reactors.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of any modifications will
be subject to the results of NRC staff review of this analysis.
TYPE OF REVIEW
Postimplementation Review.
DOCUMENTATION REQUIRED
No additional documentation is required at this time.
TECHNICAL SPECIFICATION CHANGES REQUIRED
No.
REFERENCES
Letter from D. F. Ross Jr. (NRC) to All B&W Operating Plants, dated August
21, 1979.
NUREG-0645, Volume 1, Section 2.4.6.
.
SMALL-BREAK LOCA WHICH REPRESSURIZES
THE RCS TO THE PORV SETPOINT
(II.K.2.20) POSITION
Provide an analysis which shows the plant response to a small break loss-of-
coolant accident during which the reactor coolant system is repressurized to
the PORV setpoint with subsequent failure of the PORV to close.
CLARIFICATION
The requirements was originally sent to the B&W licensees in a letter from
D. F. Ross Jr. (NRC) to All B&W Operating Plants, dated August 21, 1979.
The results of this analysis has been submitted by the B&W licensees and is
presently undergoing staff review.
APPLICABILITY
All B&W Operating Reactors.
IMPLEMENTATION DATE
Confirmatory information requested. Implementation of any modifications will
be the subject to the results of NRC staff evaluation of this analysis.
TYPE OF REVIEW
Postimplementation Review.
DOCUMENTATION REQUIRED
No additional documentation is required at this time.
TECHNICAL SPECIFICATION CHANGES REQUIRED
No.
REFERENCES
Letter from D. F. Ross Jr. (NRC) to All B&W Operating Plants, dated August
21, 1979.
NUREG-O565, Recommendation 2.6.2.c
NUREG-0645, Volume 1, Section 2.4.6
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