ML23279A071

Response to NRC Dresden Nuclear Power Station, Units 2 and 3 - Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011

ML23279A071
Person / Time
Site:	Dresden
Issue date:	10/06/2023
From:	Patrick Boyle Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, NRC/RGN-III, Document Control Desk
References
23-0032, IR 2023011
Download: ML23279A071 (1)

Text

Constellation October 6, 2023 SVPL TR #23-0032 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Dresden Nuclear Power Station 6500 North Dresden Road Morris, IL 60450 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Response to NRC Dresden Nuclear Power Station, Units 2 and 3 Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011

References:

1. Letter from Jamie C. Benjamin (U.S. Nuclear Regulatory Commission) to David Rhoades (Constellation Energy Generation, LLC), "Dresden Nuclear Power Station, Units 2 and 3 Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011," dated September 6, 2023 (ADAMS Accession No. ML23242A188)

2. Letter from Kenneth R. Riemer (U.S. Nuclear Regulatory Commission) to Bryan C. Hanson (Exelon Generation Co., LLC), "Dresden Nuclear Power Station, Units 2 and 3 - Integrated Inspection Report 05000237/2020001 and 05000249/2020001," date May 11, 2020 (ADAMS Accession No. ML20133J811)

In Reference 1, the U.S. Nuclear Regulatory Commission (NRC) identified a Green finding and associated Non-Cited Violation (NCV) during the performance of the Comprehensive Engineering Team Inspection (CETI) at the Dresden Nuclear Power Station. The finding and NCV are associated with Dresden's failure to establish measures as required by 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," to assure a condition adverse to quality was corrected. Specifically, corrective actions for a 2020 test control violation issued in Reference 2 did not account for instrument uncertainty in the accident analysis or Technical Specification (TS) surveillance requirement (SR) implementing procedure.

Issuance of this violation indicates that it is the position of the NRC that the performance of a sensitivity study for high pressure coolant injection (HPCI) flow in accordance with 10 CFR 50, Appendix K was an inadequate corrective action.

The attachment to this letter provides the Constellation Energy Generation, LLC (CEG) response and basis for the denial of the violation and finding issued to Dresden in Reference 1

U.S. Nuclear Regulatory Commission October 6, 2023 Page 2 (NCV 05000237/2023011 and 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation).

There are no regulatory commitments made within this letter. If you have any questions concerning this letter, please contact Mr. Daniel J. Murphy, Regulatory Assurance Manager, at (815) 416-2800.

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Site Vice President Dresden Nuclear Power Station

Attachment:

Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation cc:

NRC Regional Administrator, Region Ill' NRC Director, Office of Enforcement NRC Senior Resident Inspector - Dresden Nuclear Power Station

ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation In Reference 1, the Nuclear Regulatory Commission (NRC) issued a Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and associated Green finding to the Dresden Nuclear Power Station (DNPS) for the NRC's conclusion that Constellation Energy Generation, LLC (CEG) failed to correct a condition adverse to quality. Specifically, corrective actions for a 2020 test control violation issued in Reference 2 did not account for instrument uncertainty in the accident analysis or Technical Specification (TS) surveillance requirement (SR) implementing procedure.

CEG has reviewed the details of the violation provided in Reference 1 and disagrees with the NRC assessment and conclusions related to this issue. CEG has performed a detailed review of the applicable NRC regulations and NRG-endorsed industry guidance related to instrument uncertainty and 10 CFR 50, Appendix 8, Criterion XI, "Test Control," considering the DNPS licensing basis. There are no explicit requirements related to the inclusion of instrument uncertainty for Emergency Core Colling System (ECCS) flow in Criterion XI or the DNPS licensing basis. A sensitivity study demonstrated that uncertainty was implicitly accounted for in the margins of the design basis and, as a result, no corrective action was taken to account for the uncertainty elsewhere. Therefore, DNPS is not in violation for failing to correct a condition adverse to quality because no condition adverse to quality existed at the time of the Reference 2 violation. The following discussion provided the basis and justification for this position.

CEG Review of the NRC Inspection Report The Description section of Inspection Report 2023011 provides the fundamental premise for issuing the 2023 violation as follows:

The inspectors reviewed UFSAR Table 6.3-20b for HPCI LOCA (loss of coolant accident) analysis minimum flow rate and the current accident analysis, ANP-3749P, "Dresden Units 2 and 3 Atrium 10 XM LOCA Break Spectrum Analysis with Increase ADS Flow," Revision 0. The inspectors noted the HPCI assumed flowrates were both 5,000 gpm and remained unchanged since the 2020 test control violation was identified.

Since the accident analysis did not account for instrument uncertainty, the inspector verified if TS SR 3.5.1.6 implementing procedure, DOS 2300-03, Revision 119, was revised to evaluate the test results to ensure HPCI accident analysis flowrate of 5,000 gpm could be achieved. The inspectors concluded instrument uncertainty was neither accounted for in the design analysis nor the TS SR implementing procedure and therefore, AR 04331189 corrective actions did not correct the 2020 test control violation.

This statement contends that the only acceptable means of addressing the 2020 violation were to either account for instrument uncertainty in the accident analysis flow rate or in the acceptance criteria of CEG Procedure DOS 2300-03.

Similarly, the Description section of Inspection Report 2020001 provides the premise for the original 2020 violation as follows:

For the HPCI systems, the TS Surveillance Requirement (SR) 3.5.1.6 required the licensee "to verify the HPCI pump can develop a flow rate greater than or equal to 5,000 gpm against a system head corresponding to reactor pressure." The UFSAR Page 1 of 8

ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation Table 6.3-20b described the HPCI LOCA analysis value for the Minimum Rated Flow Over Range as 5,000 gpm. After conversations with the licensee, it was also established that Technical Specifications pump flows were used as inputs into the LOCA analyses without adjustments for uncertainty. This condition was repeated as the licensee transitioned through different nuclear fuel vendors (GE, Westinghouse, and Areva).

Procedure DOS 2300-03, "High Pressure Coolant Injection System Operability and Quarterly 1ST Verification Test," Revision 11, was the implementing procedure to comply with TS SR 3.5.1.6 Step H.5.a and Data Sheet 2 to this procedure established a HPCI flow of equal or greater than 5,000 gpm in order to meet the procedure's acceptance criteria. Flow was measured using Flow Controller FIC 2(3)-2340-1. In licensee's calculation NED-I-EIC-0109, Revision 6, Section 13, "Conclusions", the associated flow instrument uncertainties were determined (these vary depending on the model of the instrument installed and the unit on which it is installed). The average instrument uncertainty associated with FIC 2(3)-2340-1 was+/- 255 gpm under normal plant conditions and +/- 846 gpm under accident conditions.

Based on the above, the inspectors were concerned that failure to account for said uncertainties could result in a situation where a surveillance test was declared satisfactory when, in reality, the structure system or component (SSC) could be within the unacceptable results range once uncertainties were considered.

This statement contends that that because the TS limit, LOCA analysis input, and surveillance procedure accept criteria use the same numerical value, that instrument uncertainty has not been adequately addressed by DNPS. The wording of the premise indicates the NRC position that addressing instrument uncertainty in the surveillance procedure is the appropriate corrective action.

The original 2020 violation cites 10 CFR 50, Appendix B, Criterion XI, Test Control, as being the applicable regulation requiring a licensee to account for instrument uncertainty in the surveillance procedure; the violation did not refer to any other lower-level documents as a basis for the cited requirement. However, as shown in the next section, Criterion XI does not explicitly refer to uncertainties, but instead only requires that written test procedures incorporate the requirements and acceptance limits contained in applicable design documents. Similarly, other regulations below do not require explicit inclusion of uncertainty. Since there is no regulatory requirement that states that instrument uncertainty must be addressed in analyses or testing, the 2020 violation does not represent a condition adverse to quality.

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ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation Regulatory Requirements CEG performed a review of the applicable regulatory requirements in an attempt to understand the origin of the requirement to address measurement instrument uncertainty. The relevant regulatory basis is summarized below.

10 CFR 50, Appendix A, Criterion 13, Instrumentation and Control states:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

DNPS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC), but instead licensed to the draft GDC published in 1967. As a result, the corresponding DNPS GDC is Criterion 12, which states:

Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.

10 CFR 50, Appendix 8, Part XI, Test Control states:

A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate, proof tests prior to installation, preoperational tests, and operational tests during nuclear power plant or fuel reprocessing plant operation, of structures, systems, and components. Test procedures shall include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. Test results shall be documented and evaluated to assure that test requirements have been satisfied.

10 CFR 50, Appendix B, Part XII, Control of Measuring and Test Equipment states:

Measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.

A review of these regulations confirmed a requirement to monitor and maintain parameters within acceptance limits contained in applicable design documents. There is no prescribed reference to the treatment of instrument uncertainties beyond the maintaining the measurement and testing devices calibrated and within required accuracy limits.

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ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation Dresden Licensing Basis DNPS Units 2 and 3 went into commercial service in June 1970 and November 1971, respectively (UFSAR Section 1.1.1 ). The DNPS original plant licensing basis was developed prior to issuance of Regulatory Guide 1. 70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," before the ECCS rules were published in 10 CFR 50.46 and 10 CFR 50, Appendix K on January 4, 197 4, and prior to issuance of Regulatory Guide 1.105, "Setpoints for Safety-Related Instrumentation." As a result, instrument uncertainty is not included in the original licensing basis or the ECCS analysis basis for DNPS, Units 2 and 3.

DNPS, Units 2 and 3 were designed and licensed as nominal value plants, meaning that the inclusion of explicit allowances for instrument uncertainties in the TS limits and limiting condition for operation (LCO) values was not a requirement or common practice. The DNPS original licensing basis review used a general evaluation method. This approach for plants of the vintage of DNPS, Units 2 and 3 has long been understood by the NRC as discussed in NUREG-0138, "Staff Discussion of 15 Technical Issues Listed in Attachment to November 3, 1978, Memorandum from Director, NRR to NRR Staff," (Reference 8, Pages 13-3 and 13-4).

NUREG-0138 describes this general evaluation method as follows:

Each set point value is based upon the most limiting transient or postulated accident condition associated with the bases for that set point. The magnitude of this safety margin and the resulting set points are established to ensure that there is a low probability of the margin being removed by an adverse combination of instrument calibration error, instrument error and instrument drift.

NUREG-0138 further describes the general evaluation method as follows:

In this approach, the discrete components of each of the margins to safety in trip set point values are not evaluated on an individual basis but are included in an overall safety margin. Each set point value is based on the most limiting transient or postulated accident condition associated with the bases for that set point. The magnitude of this safety margin and the resulting set points are established to ensure that there is a low probability of the margin being removed by an adverse combination of instrument calibration error, instrument error and instrument drift. The staff believes that this method is acceptable.

NUREG-0138 goes on to discuss that the plants evaluated under the generalized method have a "trip setpoint value and 'allowable value' [that] have the same numerical value." Following this acknowledgment, the NUREG drew the following conclusion:

The staff concludes that adequate safety margins are provided by the trip set points now in use for operating plants and that this issue does not warrant revision to any existing license or any shift in priorities.

The aggregate of this discussion in NUREG-0138 shows that the NRC concluded that there is sufficient margin in the design analysis to account for instrument uncertainty based on individual analytical limit evaluations that have considered instrument uncertainties. Given that the NRC discussion in NUREG-0138 pertains specifically to instrument setpoint methodology, the Page 4 of 8

ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation rationale directly describes the degree of adherence to Regulatory Guide 1.105 required for plants licensed prior to 1977 under the general evaluation method described in the NU REG.

This approach to setpoint allowable values is transferrable to ECCS limits included in surveillance requirements which, while not setpoints, were considered in the same context given the differences in evaluation methods at the time of NUREG-0138's writing.

As required by Appendix B to Part 50-Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Dresden System Operating Surveillance Procedures ensure the operability of systems required by TS. For the ECCS systems, performance of the TS surveillance procedures verifies the flow rate and discharge pressure meet the TS acceptance criteria. This is accomplished through periodic operability testing using installed calibrated instrumentation to ensure the indicated parameters meet TS requirements. The plant designer, General Electric Company (GE), defined accuracy and range requirements for each instrument in the Design Specification Data Sheets (Reference 3).

Per UFSAR Section 6.3.2.3.2, the HPCI subsystem is designed to pump 5600 gpm into the reactor vessel across the pressure range of 1135 psia to 165 psia. The flow verification tests for the HPCI System (TS SR 3.5.1.6 and SR 3.5.1. 7) are performed at two different pressure ranges such that system capability to provide rated flow against a system head corresponding to reactor pressure is tested at both the higher and lower operating ranges of the system. The acceptance criteria for both surveillances is that the measured flow be greater than or equal to 5000 gpm. This limit is also the input utilized in the NRG-approved 10 CFR 50, Appendix K LOCA methodology.

The TS limit of 5000 gpm was originally specified by GE (Reference 15). That limit has been validated as sufficient to assure safe plant operation as part of the original approval of the DNPS, Units 2 and 3 TS as well as the transition to the Improved Standard TS (Reference 12),

and various fuel vendor transitions at DNPS that required updates to the LOCA analysis (References 13 and 14). At no time has the TS limit of 5000 gpm been challenged as being nonconservative because of measurement uncertainty concerns during NRC review and approval of these licensing actions.

Reasonable assurance of operability to provide design function during the most limiting transient or postulated accident condition is further supported by the diverse and redundant indications available to monitor pump functions. These redundant and diverse indications were intended by design to maintain an overall safety margin and are periodically tested to detect and correct drift or inaccuracies. It is improbable that any combination of instrument uncertainty could prevent the operation of a safety function or result in a consequential reduction of margin.

A related position is also captured using slightly different language in NRC Inspection Manual Part 9900 Technical Guidance (Reference 4) as follows:

The TS limits are established with allowance for measurement tolerances already incorporated. The limits take into consideration measurement uncertainties as necessary to assure safe plant operation. The stated limit presupposes that the licensees have tolerances consistent with normal industry standards (e.g., ASTM, IEEE, AC/, etc.).

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ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation For plants licensed prior to 1974 and reviewed under the general evaluation method described in NUREG-0138, the method of the measurement tolerances being "already incorporated" was by selection of the numerical value of the associated TS limit (as described in the excerpts from NUREG-0138 quoted in the section above). The uncertainty is implicitly incorporated in the design through significant safety margins built into the analyses.

This is not the first time that the NRC has challenged the application of instrument uncertainties for ECCS flow at CEG facilities licensed during the same timeframe as DNPS, Units 2 and 3. In an inspection report dated May 6, 1998 (Reference 9), the NRC opened an unresolved item (URI) for Quad Cities Nuclear Power Station, Units 1 and 2 (i.e., URI 50-254(265)98-201-03) related to accounting for instrument uncertainty in Low Pressure Coolant Injection (LPCI) and Core Spray (CS) surveillance procedures. This URI was closed in an NRC inspection report dated July 30, 1999 (Reference 10), as follows: "This unresolved item dealt with a concern over what uncertainty factors for flow measurements needed to be included. This issue is under review by the Office of Nuclear Reactor Regulation (NRR) as a generic concern. Once NRR has completed its review, the licensee will be informed of any necessary actions by separate correspondence. Therefore this item is closed."

Since no separate correspondence pertaining to this URI was received, it is apparent that NRR did not have any concerns with the manner in which ECCS uncertainties were treated at QCNPS, Units 1 and 2. The QCNPS, Units 1 and 2 approach is consistent with how ECCS flow uncertainties have been treated at DNPS, Units 2 and 3.

Conservatism of the Appendix K Methods As provided in Reference 5 in 1971, the evaluation of ECCS performance during a LOCA assured adequate safety was "obtained from an appropriately conservative analysis." This conservatism is built into the DNPS licensing basis LOCA analysis and sensitivity studies have shown ECCS flow uncertainty to not be a sensitive feature of the analysis. In Reference 6, the attributes of the 1971 evaluation were put into the features of Appendix K and in Reference 7, amendments to 10 CFR 50.46 and Appendix K "were motivated by the fact that since the promulgation of 50. 46 [...] and the acceptable and required features and models specified in Appendix K, considerable research had been performed that greatly increased the understanding of EGGS performance." In 1983, in SECY-83-472, the NRC acknowledged confirmation through the use of realistic LOCA computer codes, that there is approximately 1000°F to 1200°F margin between the peak clad temperature (PCT) expected during the limiting large break LOCA and the 10 CFR 50.46 limit of 2200°F (Reference 11 ).

Uncertainty was not discussed outside of the sources of heat, critical heat flux, and Post Critical Heat Flux Transfer Correlations until 1988 when non-Appendix K models began to be allowed.

The NRC confirmed that Appendix K methods were "highly conservative and that the actual cladding temperatures which would occur during a LOGA would be much lower than those calculated by Appendix K." As a result, 10 CFR 50.46 was revised to delete the requirement that features of Appendix K be used to develop the evaluation model. However, "if an applicant or licensee elects to use the new realistic models, they will be required to provide sufficient supporting justification to validate the model and include comparisons to experimental data and estimates for the uncertainty." "A new paragraph (ii) was added[...] to allow features of section I of Appendix K to be used as an alternative to performing the uncertainty evaluations Page 6 of 8

ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation specified in (a)(1)(i)." DNPS opted to continue using the originally licensed Appendix K model which bounds instrument uncertainty through conservatisms built into other features of the model. DNPS was not required to provide additional supporting justification to validate the model and include comparisons to experimental data and estimates for the uncertainty.

Conclusion There is no regulatory requirement to explicitly account for HPCI flow uncertainty in either the surveillance procedure or the LOCA analysis. The applicable regulatory requirement is to monitor and maintain parameters/variables within prescribed operating ranges using measurement and testing devices with specific accuracies. DNPS credits a combination of the flow measurement instrumentation accuracy specified by GE, the TS limit being specified to account for standard industry uncertainties, and the conservatism in the LOCA analysis method.

This combined margin between the ECCS analysis results and acceptance criteria ensures that the measurement uncertainties for ECCS flow instrumentation are bounded and no safety concern exists. This has been confirmed through sensitivity studies for variations in ECCS flow.

During the timeframe when DNPS, Units 2 and 3 were licensed, instrument uncertainties were not a required part of the licensing basis. Therefore, it is inappropriate to issue a finding to CEG for not implementing a corrective action to account for HPCI flow uncertainty (e.g., revise the surveillance procedure) given that this is not a requirement in the current DNPS licensing basis.

References

1. Letter from Jamie C. Benjamin (U.S. Nuclear Regulatory Commission) to David Rhoades (Constellation Energy Generation, LLC), "Dresden Nuclear Power Station, Units 2 and 3 Comprehensive Engineering Team Inspection Report 05000237/2023011 and 05000249/2023011," dated September 6, 2023 (ADAMS Accession No. ML23242A188)

2. Letter from Kenneth R. Riemer (U.S. Nuclear Regulatory Commission) to Bryan C.

Hanson (Exelon Generation Co., LLC), "Dresden Nuclear Power Station, Units 2 and 3 -

Integrated Inspection Report 05000237/2020001 and 05000249/2020001," date May 11, 2020 (ADAMS Accession No. ML20133J811)

3. General Electric Specification 257HA353AB, "High Pressure Coolant Injection System -

Data Sheets," Revision 3, dated October 21, 1969

4. NRC Inspection Manual Part 9900: Technical Guidance, STS30T.TG, Standard Technical Specifications Section 3.0 Acceptable Measurement Tolerances for Technical Specification Limits, issued October 1, 1978

5. Federal Register 36 FR 12247-12250, "Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors," dated June 29, 1971

6. Federal Register Vol. 39, No. 3, pp 1002, Part SO-Licensing of Production and Utilization Facilities Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled Nuclear Power Reactors, dated January 4, 197 4 Page 7 of 8

ATTACHMENT Basis for the denial of NCV 05000237/2023011; 05000249/2023011, Failure to Correct Condition Adverse to Quality after HPCI Test Control Violation

7. Federal Register Vol 53, No 180 pp 35996-36002, 10 CFR Part 50 Emergency Core Cooling Systems Revisions to Acceptance Criteria, dated September 16, 1988

8. NUREG-0138, "Staff Discussion of 15 Technical Issues Listed in Attachment to November 3, 1978 Memorandum from Director, NRR to NRR Staff," Issue No. 13 Instrument Trip Setpoints in Standard Technical Specifications, dated November 1976

9. Letter from J. F. Stolz (NRC) to 0. D. Kingsley (ComEd), "Quad Cities Nuclear Power Station - Design Inspection (NRC Inspection Report Nos. 50-254(265)/98-201)," dated May 6, 1998 (ADAMS Accession No. ML20247E545)

10. Letter from J. M. Jacobson (NRC) to 0. D. Kingsley (Com Ed), "NRC Inspection Report 50-254(265)/99014(DRS)," dated July 30, 1999 (ADAMS Accession Nos. ML20210M469 and ML20210M482) 11. SECY-83-472, "Emergency Core Cooling System Analysis Methods," dated November 17, 1983

12. Letter from (U.S. NRC) to Oliver Kingsley (Exelon Nuclear), "Issuance of Amendments (TAC Nos. MA8382 and MA8383)," dated March 30, 2001 (ADAMS Accession No.

M LO 12280044)

13. Request for License Amendment Regarding Transition to Westinghouse Fuel (ADAMS Accession Nos. ML060620358 and ML060620362)

14. Letter RS-23-008 from P.R. Simpson (Exelon Generation) to U.S. NRC, Request for License Amendment Regarding Transition to AREVA Fuel, dated February 6, 2015 (ADAMS Accession No. ML15055A154)

15. Draft Provisions for Emergency Core Cooling, Dresden Unit 2, dated November 14, 1966 Page 8 of 8