United States Nuclear Regulatory Commission - Protecting People and the Environment

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 
.

                                  - 3 -

                                 (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. 
.

                                  - 3 -

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

Page Last Reviewed/Updated Monday, June 17, 2013