ML050540280
| ML050540280 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/30/2003 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| FOIA/PA-2004-0277 | |
| Download: ML050540280 (6) | |
Text
EDWIN I. HATCH NUCLEAR PLANT FIRE PROTECTION INSPECTION EVALUATION AND CONCLUSION A
Introduction During performance of the TFPI at Plant Hatch, the inspectors reviewed and evaluated circuit design changes that were made to reactor pressure vessel (RPV) safety relief valves (SRVs) 21321 -FO13D and 21321 -F013G. The licensee In their Safe Shutdown Analysis Report takes credit I&f using both'SRVs to mitigate'a fire in Fire Area 2104.
-One SRV is required to be capable of being manually opened approximiately two and a half hours after the fire was started In order to ensure that the suppression pool.
temperature will not exceed the heat capacity temperature limit (HCTL) for the
.suppression pool. The licensee credits manual control of this SRV, for controlling the suppression pool heat capacity temperature limit, in order to ensure that net positive suction head will be available to the core spray pumps which are required for mitigating
.afire in this fire area. Thi, other SRV le'required to be capable of being manually
~~opened at approximately four hours after the fire was started to manually depresrz
- the RPV and achieve cold shutdown cooling conditions. In addition, nine otherS sae required to remain closed for a fire In Fire Area 2104, in order to achieve pos-frsf shutdown conditions.
B Statement of Problem In 1993 the licensee developed and Implemented a plant m'odification which Installed a:
non-safety related Rosemount 1 154GP electronic pressure transmitters on-each of the
- .four main steam lines to monitor the nuclear boiler pressure and provide backup' actuation of eleven SRVs at or near their respective mechanical set points. The backup actuation was in addition to the mechanical actuation mode of the SRVs nd ws intended to mitigate the effe cts of corrosion induced set point drift of the SRVs. The SRVs will relieve nuclear boiler pressure either by normal mechanical action or by, automatic action from an electro pneumatic control system energized from the pressure transmitters. The installed plant \\modification has two Instrument circuits, from pressure transmitters Installed on two steam lines, running in the same cable tray In Fire Area 2104 Within close proximity to each other. Neither of these instrumentation cables was protected from fire damage in accordance With the requirements of 1 0 CFR 50 Appendix R,Section II.-G.2.
A credible fire in this area will damage the cable insulation of both of these instrument
-circuits and create abnormal leakage currents. Additionally, because an analog instrument circuit transmits low level electrical signals, leakage currents caused by cable insulation damage can measurably Impair circuit performance in a manner that has functional implications. Excess leakage currents will cause the instrument loops to fail
- high which is indicative of high nuclear boiler pressure. The electro pneumatic control system will respond to this event by spuriously opening SRVs 21321-FO13D3 and 21321 F013G and defeat the capability to manually control these SRM..The nine other
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SRVs that are required to remain closed for a fire in Fire Area 2104, will also be spuriously opened by the same fire induced damage to the two Instrument circuits.
Simultaneous opening of all eleven SRVs will result in the sudden depressurization of'.:
the nuclear boiler with a loss of reactor coolant inventory, and a loss of manual control of SRVs 2B21-FO13D and 2121-F013G. The loss of the required manual control of SRVs 2B21-FOI3D and 2121-FO13G will affect the licensee's shutdown capability and, prevent a post-fire safe shutdown for a fire in Fire Area 2104. Additionally, the spurious opening of the nine other SRVs with a subsequent loss of manual control of these SRVs, will affect the licensee's shutdown capability and prevent achieving post-fire safe shutdown conditions for a fire In Fire Area 2104.
C Evaluation of Licensee's Response The licensee in their letter, Reference one, page E-3, confirmed that two instrumentation circuits and associated relay logic, could spuriously open eleven SRVs. In their response letter the licensee also stated, that the instrumentation circuits and their assocdated logic are not required in order for the SRVs to perform their safety function, and are therefore associated circuits". It is the licensee's position that the plant modification that installed the instrumentation circuits was a non-safety related design change and that DCR # 91-134 implemented the modification In a manner fully consistent with its design Input requirements.
The licensee quotes from Generic Letter 81-12 Clarification Letter: Mattson to Eisenhut, dated March 22, 1982, (Reference 2). This letter states in part that; "Associated Circuits of Concern are defined as those cables, (safety related, non-safety related, Class 1 E, and non-Class 1 E) that:
2 Have one of the following:
b a connection to circuits of equipment whose spurious operation would adversely affect the shutdown capability (Eg. RHR/RCS isolation valves, PDS valves, PORVs)
(See Diagram 2b)
Diagram 2b clearly shows a valve/pump with its circuit conductors routed through a fire area where train WA" circuit conductors are also routed.
The licensee also referred to NRC memorandum from John A. Hannon to Gary M.
Holahan dated November 29, 2000, concerning clarification of the term, " an associated circuit". The description of an associated circuit contained in this memorandum, and the requirements in Inspection Procedure 71111.05, for not inspecting associated circuits as a direct line of inquiry was reiterated by the licensee. The licensee concluded that, based on NRC present guidance, the concern expressed by the inspectors involving plant modification DCR # 91-134 was beyond the design and licensing bases of the plant.
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D Post-Fire Safe Shutdown Equipment The licensee provided reference three in response to a request for additional information. Figure 1 documented in reference three shows the modified control /
protection circuit for SRV 2B21-F013B. This circuit is typical of the circuits installed for the following four SRVs that the licensee credits for mitigating a fire in any area in the plant. The post-fire safe shutdown equipment addressed in this evaluation are those required for Shutdown Path No. 2.
Shutdown Path No. 1 SRVs Shutdown Path No. 2 SRVs 2B21 -FOl 3B 2B21 -FOl 3D 2B21 -FO1 3F 2B21-FO13G Path one post fire safe shutdown equipment consists of the reactor core isolation cooling (RCIC) system and the SRVs shown which are used to provide reactor protection via depressurization, inventory makeup, and decay heat removal. Path two post fire safe shutdown equipment consists of the high pressure coolant injection (HPCI) system and the SRVs shown which are used to provide reactor protection through depressurization, inventory makeup, and decay heat removal. All four SRVs are required post-fire safe shutdown equipment.
E Plant Licensing and Design Bases Review E.1 The licensee asserts that the two 4-20 milli-amp Instrumentation circuits are non-safety related circuits, as shown in plant modification DCR# 91-134.
The licensee is committed to IEEE Standard 279-1971, as documented in HNP-FSAR, section 8.1.4.H.3, Institute of Electrical and Electronics Engineers (IEEE) Standards.
Section 4.7 of this standard describes the requirements for classification of equipment that is used for both control and protective functions. The SRVs are used to control the suppression pool heat capacity temperature limit (HCTL) and to provide over-pressure protection for the nuclear boiler. The 4-20 milli-amp instrumentation circuits are part of the Analog Transmitter Trip System (ATTS) that transmits the nuclear boiler pressure signals for actuating the relay logic in order to provide the over - pressure protective function. In accordance with the guidance given in IEEE Std. 279, section 4.7.1, Classification of Equipment, these circuits are required to be classified as part of the protection system. The 4-20 milli-amp instrumentation circuits are therefore Class 1 E instrumentation circuits which is consistent with the requirements of the licensee's FSAR, Section 7.1.2.1, Design Bases, where one of the generating station variable that require monitoring to provide protective actions include " RPV Pressure".
E.2 The licensee asserts that the two conductor control cable routed from control panel 2H1 1-P927 to control panel 2H1 1-P925 is a non-safety related circuit, as shown in plant modification DCR# 91-134.
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The licensee is committed to IEEE Standard 308-1971, as documented In HNP-FSAR, section 8.1.4.H.3, Institute of Electrical and Electronics Engineers (IEEE) Standards.
This standard was revised in 1978 to clarify the connection of non-Class 1 E loads to Class I E buses. Section 5.10 of this standard describes these requirements and endorses the guidance delineated in IEEE Standard 384-1977 concerning this subject The purpose of IEEE Standard 384-1977 was to establish the criteria for implementing the independence requirements of IEEE Standard 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations.
The power source for the circuit shown on Figure I is from the 125 VDC Class 1 E Station Batteries, which classifies this circuit as a 125 VDC Class 1 E control circuit. The general criteria for interconnection of non-Class 1 E circuits with Class 1 E circuits are given in IEEE Std. 384-1977. A review of IEEE 384, sections 4.6 and 6.2, does not justify the licensee's classification of the control cable routed between both control panels. The specific electrical isolation criteria delineated in this standard for non-Class 1 E circuits are not applicable to Figure 1. Additionally, the definition of an electrical circuit given in IEEE Std. 100 -1977, defines an electrical circuit as (1) a network' providing one or more cdosed paths or (2) an intercon..c^ion of electrical elements.
The licensee's statement that the two conductor control cable is a non-safety related circuit is therefore Incorrect.
The power source for this circuit is from the 125 VDC Class 1 E Station Batteries.
Additionally, the isolation fuses In control panel 2H1 -P925 are intended to prevent the propagation of electrical faults, originating in control panels 2H11 -P927 or 2H 1 -P928 because of equipment failure, from degrading the Class 1 E control circuit. This design featureis-used-when-a-non-Class -E-non-safety-related-load-is-wiredinto-a-Olass1 E circuit. Using the guidance delineated In the IEEE standards, the control circuit shown on Figure 1 should be considered a safety related, 125 VDC Class I E control circuit.
The electrical conductors, ie cables, for this circuit are routed from the 125 VDC power source up to, and including all conductors In control panels 2H1 1 -P927 and 2H1 1 -P928.
E.3 The licensee asserts that the two conductor control cable routed from control panel 2H1-1-P927 to control panel 2H1-1-P925 and the two 4-20 mill-amp Instrumentation circuit cables are associated circuits.
The licensee's statement Is based on the definition of an associated circuit given in Letter 81-12 Clarification Letter: Mattson to Eisenhut, dated March 22, 1982," Fire Protection Rule-Appendix R", and Diagram 2b which is part of the clarification. A review of Figure 1 and comparison with Diagram 2b shown in the generic Letter 81-12 Clarification Letter, reveals significant differences. Diagram 2b clearly shows a valve/pump with its circuit conductors routed through a fire area where train "A" circuit conductors are also routed. The licensee's classification for an associated circuit/non-safety related circuit is applied erroneously to specific cables, which are part of the electrical circuit. This has resulted In a situation where the electrical circuits of post-fire safe shutdown equipment, SRVs 2B21-F013D and 2B21-F013G, will have a circuit conductor classified as either a required circuit or an associated circuit, depending on which cable in the circuit one Is addressing. This is contrary to accepted industry 4
C practice and electrical engineering principles. It is also contrary to the definition of an electrical circuit as defined in IEE Std. 100 -1977, which defines an electrical circuit as (1) a network providing one or more closed paths or (2) an interconnection of electrical elements.
F Conclusion Based on review and evaluation of the licensee's response transmitted In their letter dated October 1, 2003 to Inspection Report 50-321/50-366 2003006, and additional information e-mailed to Region II office, the NRC has concluded the following:
(1)
The licensee's assertions, delineated in Section E of this Evaluation' and Conclusion, do not satisfy the requirements of 10 CFR 50, Appendix R, Section III.G.2, and the plants licensing basis as demonstrated by commitments to IEEE Standard 279-1971, and FSAR, Section 7.1.2.1, Design Bases.
(2)
Accordingly, the licensee was required to provide physical protection for the two instrumentation circuits in accordance with the requirements of 10 CFR 50 Appendix R, Section III.G.2, to ensure that post-fire safe shutdownbconditions were achieved using the SRVs that are credited with mitigating a fire in Fire Area 2104. The operator action placed in the fire procedure to prevent spurious opening of SRVs 2B21-F013D and 2621-F013G, in the event of fire induced damage to the two instrumentation cables in Fire Area 2104, is considered manual compensatory actions, not approved by the NRC, that was implemented to satisfy the requirements of 10 CFR 50 Appendix R, paragraph IIl.G.2.
(3)
Unresolved item, URI 50-366103-06-02, Untimely and Unapproved Manual Action for Post-Fire SSD, correctly describes this manual compensatory action and remains open pending completion of a significance determination.
(4)
The spurious openings of nine SRVs that are required to remain closed during a fire In Fire Area 2104, are identified as an associated circuit Issue that arose unavoidably during the inspector's review of safe shutdown system equipment for mitigating a fire in Fire Area 2104. In accordance with the guidance of IP 71111.05 this byproduct associated circuit Issue is documented as URI 366/03-06-01, Concerns Associated with Potential Opening of SRVs. Also, this URI will be kept open pending generic resolution of the related associated circuit issues, in accordance with the guidance delineated in IP 71111.05 (5)
Unresolved Item, URI 50-366/03-06-06, Inspector Concerns Associated with Implementation of DCR 91-134, will be closed based on review of additional information provided by the licensee.
Reference Documents 5
- 9.
I (1)
Southern Nuclear Operating Companies (SNC) letter dated October 1, 2003,
Subject:
Edwin I. Hatch Nuclear Plant, Response to Inspection report 50-321/50-366 2003006 (2)
Roger J. Mattson letter to DarreIG. Eisenhut, dated March 22,1982,
Subject:
Fire Protection Rule - Appendix R (3)
Additional Information provided by SNC, Use of the Term, One out of two taken twice."
(4)
IEEE Standard 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations".
(5)
IEEE Standard 308-1971, "Criteria for Class 1 E Electric Systems for Nuclear Power Generating Stations".
(6)
IEEE Standard 100-1977," IEEE Standard Dictionary of Electrical and Electronic Terms".
(7)
IEEE Standard 384-1977, U IEEE Standard Criteria for Independence c. Lass E Equipment and Circuits".
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