ML20209H299
| ML20209H299 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 07/16/1999 |
| From: | Krich R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9907200183 | |
| Download: ML20209H299 (5) | |
Text
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Commonwealth Edhon Company 124KI Opus Place w
Downers Grove. IL N1515-5 701 July 16,1999 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Facil;ty Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Response to Request for Additional Information Regarding License Amendment Request to Credit Automatic Power-Operated Relief Valve l
(PORV) Operation for Mitigation of inadvertent Safety injection at Power Accident and Withdrawal of License Amendment Request l
References:
(1)
Letter from K. L. Graesser (Comed) to U. S. NRC Document l
Control Desk, " Change to Credit Automatic PORV Operation for Mitigation of inadvertent Safety injection at Power Accident," dated May 29,1998.
(2)
NRC le+ter, " Request for Additional Information - Byron Station, i
Units 1 and 2 and Braidwood Station, Units 1 and 2," dated May 13,1999.
A License Amendment Request (LAR) for Byron Station and Braidwood Station to credit automatic operation of the pressurizer Power-Operated Relief Valves (PORVs) to mitigate
' a spurious safety injection (SI) at power accident was submitted to the NRC in Reference 1.
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9907200183 990716
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PDR ADOCK 05000454 p
PDR 4UUUbb A Unicum Company
July 16,1999
> U. S. Nuclear Regu'atory Commission Page 2 The NRC issued a Request for Additional Information (RAl) letter (Reference 2) regarding the instrumentation circuitry for automatic operation of the pressurizer PORVs. The instrumentation circuitry for automatic operation of the pressurizer PORVs does not meet the requirements set forth in the Institute of Electrical and Electronics Engineers (IEEE)
Standard 279 - 1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," as required by 10 CFR 50.55a, " Codes and standards." The RAI letter
. requested that additional information be provided within 60 days after receipt of the letter
. (i.e., by July 16,1999). The requested additional information is provided in the attachment to this lettar As notW N ihe response to Request 1, we are continuing to evaluate methods to address the automatic operation of the pressurizer PORVe to mitigate a spurious Si at power accident. Due to the ongoing nature of these evaluations, we are withdrawing, from NRC consideration, the LAR submitted in Reference 1. If a change is deemed the appropriate resolution for this issue, another LArs am oe provided. Accordingly, an affidavit is attached.
Should you have any questions concerning this letter, please contact Ms. K. M. Root at (630).663-7292.
Respectfully,.
R. M. Kri[,
Vice President - Regulatory Services Attachments cc:
Regional Administrator-NRC Region til NRC Senior Resident inspector - Braidwood Station NRC Senior Resident inspector-Byron Station t
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July 16,1999 U. S. Nuclear Regulatory Commission Page 3 i
ATTACHMENT i
STATE OF ILLINOIS
)
j COUNTY OF DUPAGE
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IN THE MATTER OF
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COMMONWEALTH EDISON (COMED) COMPANY
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Docket Numbers j
BRAIDWOOD STATION UNITS 1 AND 2
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STN 50-456 AND STN 50-457 BYRON STATION UNITS 1 AND 2
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STN 50-454 AND STN 50-455
SUBJECT:
Response to Request for Additional information Regarding License Amendment Request to Credit Automatic Power-Operated Relief Valve (PORV) Operation for Mitigation of Inadvertent Safety injection at Power Accident and Withdrawal of License Amendment Request AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information and belief.
M OFFICIAL SEAL LINDA S. SANDERS Vice President - Regulatory Services f N0lWPUBUC.SWE OFIWNOIS MCOWWISSION EXPlRES 12 29 200{
Subscribed and swom to before me, a Notary Public in and
/U/d for the State above named, this
_ day of M t/
.19 N.
Notary Public
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ATTACHMENT Response to Request for Additional Information Regarding Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Change to Credit Automatic Power-Operated Relief Valve (PORV)
Operation for Mitigation of inadvertent Safety injection at Power Accident
)
Reauest 1 The following documents the bar,is for the staff's concerns that the Byron /Braidwood design does not meet the required single failure criterion.
On November 11,1998, Comed provided a set of seven drawings (6E-1-4031RY32,01, 15, 26,16,4030RY13,17] in which the class 1E portion of pressure transmitter 1PT-0455 signal to the PORV solenoid 1RY455A was shown in yellow and the non-1E portion was shown in orange. The circuit shown by these two colors indicates that the safety signal from the class 1E transmitter and the isolation relay goes through a series of non-1E devices including relay PY455EX, whose output contact is considered class 1E, to initiate an automatic actuation of the associated PORV solenoid. An additional set of two drawings [6E-1-4031RYO4,13] for the signal from pressure transmitter 1PT-0458 was also provided with similar color scheme. This set, however, does not indicate actuation of a PORV solenoid.
The instrumentation circuit for an automatic operation of a PORV to mitigate the consequences of an inadvertent safety injection (SI) should meet the requirements set forth in IEEE-279 as required by 10 CFR 50.55a for protection systems. IEEE-279 explains design requirements for control and protection system interaction in section 4.7 and the single failure criterion in section 4.2. As per section 4.7.2 of IEEE-279, an i
isolation device is used to transmit a signal from protective system equipment for control system use such that no credible failure at the output of the isolation device, i.e., no failure or fault in the non-1E portion of the instrumentation circuit, shall affect the protective function of the associated class 1E system.
The isolation devices shown on the submittal drawings do not perform the functio..
identified in section 4.7.2 of IEEE-279 and failure or a fault in any one of the several non-1E devices in the non-1E portion of the circuit will prevent automatic actuation of the
- PORV solenoid.~ Therefore, the circuit initiating automatic actuation of the PORV solenoid should, in its entirety, meet the single failure criterion of IEEE-279. Please address this issue. Incidently [ incidentally), the Salem Generating Station design involved similar problems of not meeting the single failure criterion in the automatic actuation circuit of its PORVs. The licensee incorporated several design modifications to meet the regulations.
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ATTACHMENT Response to Request for Additional Information Regarding Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Change to Credit Automatic Power-Operated Relief Valve (PORV)
Operation for Mitigation of inadvertent Safety injection at Power Accident
Response
We acknowledge that the instrumentation circuitry for automatic operation of the pressurizer Power-Operated Relief Valves (PORVs) does not meet the recommendations set forth in the institute of Electrical and Electronics Engineers Standard (IEEE) - 279, " Criteria for Protection Systems for Nuclear Power Generating Stations," as required by 10 CFR 50.55a, " Codes and standards." However, as noted by the NRC in NUREG-1316, " Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nuclear Power Plants," it would not be cost justified to backfit the recommendations for safety grade designs on non-safety-grade PORVs, block vnives, and associated control systems. Alternatively, actions were recommended to in prove the reliability of existing designs. These findings were issued in NRC Generic Letter (GL) 90-06, " Resolution of Generic issue 70, Power Operated Relief Valve and Block Valve Reliability, and Generic Issue 94, Additional Low-Temperature Overpressure Protection for Light-Water Reactors." Byron Station and Braidwood Etation have conformed to the recommendations of GL 90-06, and revised the Tech *al Specifications (TSs) accordingly. These revisions were issued in Amendment Number 44 for Byron Station and Amendment Number 33 for Braidwood Station. As noted in the NRC Safety Evaluation associated with these TS Amendments, adoption of these provisions enhances the reliability of the PORVs and bicck valves and represents a substantial increase in the overall protection of the public health and safety. As a result, it was not considered necessary to modify the instrumentation circuitry for automatic operation of the PORVs to meet IEEE-279 recommendations.
1 We are continuing to evaluate methods to address this concern. Due to the ongoing nature of these evaluations, we are withdrawing from NRC consideration the license amendment request (LAR) for Byron Station and Braidwood Station associated with crediting automatic operation of the pressurizer PORVs to mitigate a spurious safety injection at power accident. If a license amendment is deemed the appropriate resolution for this issue, another LAR will be provided.
ATTACHMENT Response to Request for Additional Information Regarding Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Change to Credit Automatic Power-Operated Relief Valve (PORV)
Operation for Mitigation of inadvertent Safety injection at Power Accident
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l Reauest 2 An additionalissue, unrelated to the single failure issue, concerns a data plot provided on November 18,1998. Specifically, Figure 4, which plotted Pressurizer Water Volume, shows a high value of approximately 1870 ft. However, Table 5.4-9 in the Updated Final Safety Analysis Report (UFSAR) documents the internal volume of the pressurizer as 1800 ft'. This discrepancy will need to be resolved.
ResDonse i
Figure 4, Pressurizer Water Volume, which we provided to the NRC by telecopy on November 18,1998, was generated using the LOFTRAN computer code. This code simulates the neutron kinatics, the Reactor Coolant System, the pressurizer, the pressurizer PORVs and safety valves, the pressurizer spray, the Feedwater System, the I
steam generator, and the steam generator safety valves, and computes pertinent plant varisbles.
The pressurizer water volume in Figure 4 was generated using the LOFTRAN computer code that accounts for pressurizer surge volume. Pressurizer surge volume is approximately 70 ft'. Therefore, an internal pressurizer volume of 1800 ft in addition to a pressurizer surge volume of approximately 70 ft' yields the pressurizer water volume of approximately 1870 ft as represented in Figure 4.
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