ML050540279

E-Mail from C. Smith to C. Ogle, Et. Al., Request for Peer Review

ML050540279
Person / Time
Site:	Hatch
Issue date:	11/21/2003
From:	Casey Smith NRC/RGN-II
To:	Fillion P, Merriweather N, O'Donohue K, Ogle C, Payne D NRC/RGN-II
References
FOIA/PA-2004-0277
Download: ML050540279 (7)

Text

Unaries H. ugle -_jeuest tor Peer Heview _ ___ age I

, . H From: Caswell Smith] 2 aiŽ To: Charles Ogle; Charlie Payne; Kathleen O'Donohue; Norman Merriweather; Paul; z2z Fillion Date: 11/21/03 11:34AM

Subject:

Request for Peer Review Attached is my evaluation of the licensee's response to the inspection findings for the TFPI that was performed at Hatch. I am requesting a critical review of my conclusions prior to sending this document to NRR for their review and concurrence. I have hard copies of the IEEE Standards referenced in the evaluation and the licensee's response letter. Thanks.

V"Ali V

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) 2B21 -F01 3D and 2B21-F013G. The licensee in their Safe Shutdown Analysis Report takes credit for using both SRVs to mitigate a fire in Fire Area 2104.

One SRV is required to be capable of being manually opened approximately two and a 2 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 a fire in this fire area. The other SRV is required to be capable of being manually opened at approximately four hours after the fire was started to manually depressurize the RPV and achieve cold shutdown cooling conditions. In addition, nine other SRVs are required to remain closed for a fire in Fire Area 2104, in order to achieve post -fire safe shutdown conditions.

B Statement of Problem In 1993 the licensee developed and implemented a plant modification which installed a non-safety related Rosemount 11.54GP 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 and was intended to mitigate the effects 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 10 CFR 50 Appendix R, Section III.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 2B21-F013D and 2B21 F013G and defeat the capability to manually control these SRVs. The nine other 1

.; . X.

SRVs that are required to remain closed for a fire in Fire Area2104, 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-F013D and 2B21-FO13G. The loss of the required manual control of SRVs 2B21-F013D and 2B21-F013G 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 associated 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; aAssociated Circuits of Concern are defined as those cables, ( safety related, non-safety related, Class 1E, and non-Class 1E) 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 "A" 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.

2

4, .4...,.4 1 FreSae hudown Euoet1/2~c2W'.,

Nin The'licensee provided reference three in response to a request fr additional-

.t ,t,'j> Informaton. Figure1 'doh'ume'nted in reference threeshows the. modified controi.. ,

'tit., Aptotetiocircuit for SRV 2B21-F013B. This-circuit is typica ofthe'ircutsinsalle for

.:,.

• th. folldwng oI4 four SRVs that the'licensee credits for mitigating a fire in any area in the -

f.plant. po'stfiresafeshutdown'equiprenttaddressedin this evaluation arethose

4 S P or, No 1 SRVs _

s'-3,', 4 Pa&No

'f 4. 2 SoRV. asho'wnre 2B21.n;--fO13 pc uip ad2B21fo13D d'in 1s att -  ;

4 t

!..- . .The p safe shutdown equi nt consistsof the reactor core isolatione -

< '.-' po'st-fO1re saf 2Btow'eui'en~onss'F03 thehi.pesue a ii.e-o (P - t~-+

',;*.` -- cooling it (ROIC) syitem adhRVs 9which shown are reactor"

/.t s, in }nos dcy at, - Path psire sateib t e high pressure clatad injetion (HPCI)

' s' stand apdthaRs Ir ,i a

-Icdnnsofeo sh.' .-..-

f i9i8tso which' 4 tŽ~,.~' cithprovide dj. reator'Jr_ poteCtion

. i+ .:v~

through -

<-,. u., depewssuzBationes entrmkwhpichiey a. at5:V removalsAl fu SRus ar

?43/4-recuited jiifr~~es'hutdo-"whtn'equipmtieniC -.

-- E Plant Licensin i DnESid Bs es Revieb a elca c control cable roiedsfrom controlpanel r EL MTe licenseese.Se assaerts' tio the.twovcoriductor that '.

th;'-'s' 6~,-i.;"Ca~ .i of ihlis-st nda'do scribetheir uitciremnts-l'l' h

. .~ . *t conrol anel2H11P925isja't n~on ;safety related circuit as showni ,

44helieneeiscomite t rui IEEStndid30- 971; as doumentedi

.W., ,-;;-,,, -p.co.rne -",-denperescntnza

,---.. nsee4 ior~n; cnveftioof,no-,Cas withb Clas .are,SAs

'A I frut i~ Ni r QN

rt- )1 P-F-SAR,

.> t^>, '=!..,"- -~.'ie in IEE

-fo4;? - Std 38 -97 . ,4 ,,

,. - ,' ,, , ,.4

.1<, .4-4.44 KtK- - sec~tion 8.1A4.H3

.. lntitut6 of Electriihl ajid Eiectr~hicsEneiriedis(lEEE) Standairds. .,,-

e ..-

.0This'stadtd 'wa's re'v'isb-d In1'978 to blariy-h

..-....t i 8ag.-z I6iecino nonCls1E!cdto7 ses E Cla.4 o ,hisstnd dd6cribesttese .requirements. .also I; t t'.17.'!.

'din"4dii~~ti a34-97concerning tissujc.,_

,The powrsuc o tl ici hwn 6n'F igure u,s -fromthea125 YDCQClass I .1..

1.. Sttio Bat~is~wih clasi~'6ifi~es this "c~i'r-litas"-a? 125 VDC'Clarss 1E &nitrolctiricth

~,-; given in IEEE Std384-i 977.eve IEE64 ttfT 8nlwosoi9-1 4

"~<

1UStr4 &wilessf'atrl¶ftD abtROUttwgfoh! nt pael Teseiieltncal isolation cnteia delfinate in ths sadrfonon-Clas 1ircitsare E not applicable to Fiue1Adtoalthdfntonfaneetca V 4 'cicu~t gven inIEEE Sd 100 977 defines an electnical circuit as' (1)anewr,

.4 .poiing on1o66cosd -paths or -(2) anin -16tncal e~~i~',;

oroiki~'f Te licesesstteet ht heto oductorco-n-tr-ol c-a-ble is a-n-on-sft eae 1 44 ciciis~theref-or-e-in;cor-r~e~ct'  ;.-<~

r-.

\.4

.4' ..

I The power source for this circuit is from the 125 VDC Class 1E Station Batteries.

Additionally, the isolation fuses in control panel 2H1-1 -P925 are intended to prevent the propagation of electrical faults, originating in control panels 2H11-P927 or 2H1-1-P928 because of equipment failure, from degrading the Class 1E control circuit. This design feature is used when a non-Class 1E / non-safety related load is wired into a Class 1E 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 1E control circuit. ~1 fJI.

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.2 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 TriD System (ATTS) that transmits the nuclear boiler pressure-signals for dctuating the relay logic in order to provide the ovPr - prr tp tive

V.),-X 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 pFoff the I'V-1 5./

protection system. The 4-20 milli-amp instrumentation circuits are therefore Class 1E /V. .e -

instrumentation circuits which is consistent with the classification of the relay logic circuits that actuates the Class 1E, safety related SRVs. This classification is also C,0 1A1_ d- allt 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.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 r

circuit cablesare associated circuits; 416- -

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 Diagramm2b.showninthe-generic-Letter-.81A2.

Clarification Letter, reveals significant differences. Diagram 2b clearly shows a valve/pump with its circuit coiidators 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 26211-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 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 4

(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 shutdown conditions 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 2B21-FO13G, 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 III.G.2.

(3) Unresolved item, URI 50-366/03-06-02, 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 (0f 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 5

(7{ Roger J. Mattson letter to Darrel G. Eisenhut, dated March 22, 1982,

Subject:

Fire Protection Rule - Appendix R (9< Additional Information provided by SNC, " Use of the Term, One out of two taken twice."

(9 IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations".

(19) IEEE Standard 308-1971, " Criteria for Class 1E Electric Systems for Nuclear Power

/ Generating Stations".

(1,5 IEEE Standard 100-1977," IEEE Standard Dictionary of Electrical and Electronic Termse.

($) IEEE Standard 384-1977, " IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits".

6