ML20196L492

Forwards Response to NRC 880217 Request for Addl Info Re NUREG-0737,Item II.D.1, Performance Testing of Relief & Safety Valves. Boeing Computer Svcs Memoranda Also Encl

ML20196L492
Person / Time
Site:	Fort Calhoun
Issue date:	06/28/1988
From:	Andrews R OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM LIC-88-477, NUDOCS 8807070532
Download: ML20196L492 (12)

Text

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1 Omaha Public Power District I 1623 Harney Omaha. Nebraska 68102 2247 l 402.536 4000 June 28, 1988 LIC-88-477 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station Pl-137 Washington, DC 20555

References:

1. Docket No. 50-285

2. Letter from NRC (A. Bournia) to OPPD (R. L. Andrews) dated February 17, 1988

3. Letter from OPPD (R. L. Andrews) to NRC (Document Contrcl Desk) dated May 27, 1938 (LIC-88-384)

Gentlemen:

SUBJECT:

Response to Request for Additional Information concerning NUREG-0737, Item II.D.1 The Omaha Public Power District (0 PPD) received Reference 2 which detailed the NRC staff and its consultant's review of NUREG-0737 Item II.D.1, Performance Testing of Relief and Safety Valves for Fort Calhoun Station.

Reference 3 was OPPD's response to the questions listed in Reference 2. The response to Question 12 indicated that additional time would be required to fully respond to the question and that OPPD's response would be submitted by June 30, 1988. Attached please find our response to Question 12. Also attached please find copies of Boeing Computer Services memoranda which were not included in Reference 3.

If you have any questions, do not hesitate to contact us.

Sincerely, AV,5,y .Nbn

,< c R. L. Andrews Division Manager Nuclear Production Attachments 8807070532 880628 d RLA/me PDR ADOCK 05000285 Ol(

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l c: LeBoeuf, Lamb, Leiby & MacRae p g R. D. Martin, NRC Regional Administrator  ;

P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector

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ATTACHMENT NRC Question 12:

NUREG-0737, Item 11.0.1 requires that the plant specific PORV Control Circuitry be qualified for design-basis transients and accidents. OPPD's response to this was, "The control circuitry for the PORV is, for the most part, located outside of the containment building, in the switchgear and control rooms. As such, it would not be subjected to a harsh environment.

The solenoid valves which open the PORVs are located at the PORVs inside containment. For the Fort Calhoun Station, the transients which might challenge the PORVs, namely loss of load or loss of feedwater flow, do not create a harsh environment in the containment. In the highly unlikely event that both PORVs failed to open when challenged, either of the two safety valves could provide more than enough capacity to handle the amount of steam that would be generated."

The licensee's statement is considered evasive since it does not address the pertinent requirements of NUREG-0737, Item II.D.1, namely, accidents and transients inside the containment that subject the PORV circuitry to harsh environment during which the PORY may operate.

The staff has agreed that meeting the licensing requirements of 10 CFR 50.49 for this circuitry is satisfactory and that specific testing per NUREG-0737 requirement is not required. Therefore verify whether the PORV control circuitry has been reviewed and accepted under the requirements of 10 CFR 50.49.

If the PORV circuitry has not been qualified to the requirements of 10 CFR 50.49, provide information to demonstrate that the control circuitry is qualified per the guidance provided in Reg. Guide 1.89, Revision 1, Appendix E.

As an alternative, the staff has determined that the requirements of NUREG-0737 regarding the qualification of the PORV control circuitry may be satisfied if one or more of the following conditions is met.

NRC POSITION 12a. The PORVs are not required to perform a safety function to mitigate the effects of any design basis event in the harsh environment and failure in the harsh environment will not adversely impact safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifically archibit use of PORVs in accident mitigation, it must be ascertained t1at PORVs can be closed under harsh environment conditions).

OPPD RESPONSE i

The following discussion provides OPPD's analysis of the expected PORV operations and anticipated plant response, during various plant operating modes.

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l.0 NORMAL OPERATION (RCS > 1700 PSIA)

For the purposes of the PORV review "normal operation" is considered to be above 1700 psia where Pressurizer low-Low Pressure (PPLS) is enabled (unblocked) and Low Temperature Over-Pressure Protection (LTOP) is thus disabled. For the normal operating condition, only the loss of load or the power increase events are applicable. In both cases the Pressurizer Quench Tank (PQT) is sized adequately to contain the primary system volume released, thus no harsh environment is created. Should one or both PORVs stick open the RCS transient which would occar is bounded by the LOCA analysis. A LOCA qualified acoustic monitoring system for PORV flow is available to insure the solenoid limit switches do not mislead the opera-tor as to the PORV's open or closed status. Please note that per OPPD's EEQ Program the PORV controls are not identified as being LOCA qualified while the PORV acoustic flow monitor is identified as being qualified via an orange dot on the control board. This pernits the operator to easily identify qualified and thus reliable instrumentation.

2.0 HEATVP AND COOLDOWN LTOP OPERATION (RCS <1700 PSIA)

During heatup and cooldown (RCS 11700 psia) the PPLS circuit is blocked and the LTOP circuit is enabled, in the event of a pressure excursion, per the LTOP circuit setpoints, the PORVs serve to reduce the RCS pressure. Under this condition the volume of RCS discharge is not speci-fically known; however, OPPD believes it to be loss than the volume of discharge as a result of an overpressure condition under "normal opera-tion". The PQT system is designed to contain the entire RCS discharge resultant from a full power loss of load trip. Therefere, the PQT system would remain intact for an overpressure condition under LTOP operation and no harsh environments would be created. This is considered conservative because of the significant difference in the energy within the RCS under the two conditions. Under "normal operation" the RCS pressure would be decreasing from 2100 psia, the cold leg temperature would be 535'F or greater, and decay heat would be exponentially decreasing from 100% reac-tor power. The conditions for which the LTOP system is enabled would be RCS fluid pressure of 1700 psia or less at approximately 450*F with decay heat below 1% reactor power. The reactor coolant pump heat would be the only other major source of heat input to the system for both scenarios.

Therefore, any LTOP transient would not be as severe , an RPS high pressurizer pressure - PORV transient.

The heat removal characteristic of the steam generators is r,ot always  ;

obylous in the comparison of RPS PORV function versus the LTOP PORV func-tion. The main steam safety valves and steam dump and bypest valves are assumed to function properly for decay heat removal in the case of high pressurizer pressure. In the case of a heatup or cooldown ovarpressure, the RCS pressure, temperature, and decay heat removal are controlled by manual control of steam generator steaming. The recovery from either sit-uation would require the use of ths: steam generator (s) for heat removal.

The existing LTOP configurnico is adequat.? and the POT would be expected

, to remain intact. Shculd C W Vs opeb ars the PQT rupture disc burst, operator action can be tal r+rol E'.S nressere and close the PORV block valves. Should the . .

to cusa the event is r,till bounded by the LOCA analysis wit! a indicat un available for the PORVs.

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3.0 NORMAL OPERATION CONFIGURATION - LOCA RESPONSE The LOCA response of the LTOP circuit due to temperature input failure would generate a PORV open signal. This would be 3revented during power operation by PPLS being unblocked which disables tie control capability of the LTOP circuit.

The area of concern here is the post-LOCA action where engineered safeguards are reset. PPLS must be blocked as the first step to reset safepiiards, which could automatically open the PORVs. OPPD will add a step w A0P-23, Safeguards Reset Procedure, to clearly require the PORV control switches to be placed in "close' position prior to blocking PPLS.

The control circuitry required to prevent energization of the PORVs would e not be subjected to a potentially harsh environment, i 1 4.0 LTOP OPERATION CONFIGURATION - LOCA RESPONSE In the event of a LOCA occurring when the LTOP is enabled, spurious actuation could occur. It is judged that the existing configuration is adequate for the following reasons:

a. Operating time with the RCS below 1700 psia and above 300 psia is limited to approximately one to two heatup and cooldown cycles per year. Each heatup or cooldown process has a duration of approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> which greatly reduces the probability of a concurrent LOCA.

b. One of the first cNrator actions following a LOCA is to unblock PPLS to initiate safeguards. This deenergizes the PORV solenoids; unless

mechanical binding occurs, the solenoids should reposition. The

solenoids are large masses of copper and iron which generate heat when energized and thus would not be greatly influenced by initial stages of LOCA induced transients. Exposure to steam heating in the first t moments of a LOCA is not expected to cause binding, t

c. Any PORV failure would be bounded by the LOCA analysis and qualified position indication is provided to prevent operator confusion.

5.0 PORV SOLEN 0ID llMIT SWITCH & S0LEN0ID ,

5.1 The PORV solenoid limit switch failure is bounded by the LOCA qualified acoustic flow position indication position indication

, discussion in Section 1.

) 5.2 The PORV solenoid failure in a harsh environment is considered bounded

! by the LOCA analysis in the case of an open PORV. A failure to open  ;

analysis is not required, as PORV opening in a harsh environment is ,

not required.

l 6.0 POST LOCA LONG TERM CORE COOLING VSING THE PORVs

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6.1 The LOCA analysis states that the PORVs can be used for long term core cooling in the event both steam generators are not available. Both steam generators would be unavailable only in the event all feedwater (main and auxiliary) was lost. The AFW system is considered to meet i

single failure criteria for events requiring decay heat removal and is of adequate reliability, thus the total loss of feedwater is not considered credible. In addition, OPPD has committed to the addition of a third AFW pump to further improve system reliability.

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7.0 ONCE THROUGH COOLING USING THE PORVs 7.1 Once through cooling of the core is required only in the event of a loss of all feedwater (main and auxiliary). In this mode the PORVs would be opened and HPSI pumps would be used to provide make-up to the RCS. This cooling mode is discussed only in E0P-20 Success Path HR 4 and is not part of the USAR 14.10 Malfunctions of the feedwater System Analysis.

Once through cooling is not considered as part of the Fort Calhoun design basis. The auxiliary feedwater system is considered of adequate reliability and meets Fort Calhoun Station design basis single failure criteria for events requiring decay heat removal.

NRC POSITION 12b. The PORVs are required to perform a safety function to mitigate the effects of a specific event, but are not scSjected to a harsh environment as a result of that event.

OPPD RESPONSE The PORVs function is discussed in the response to 12a, see Sections 1.0, 2.0, 3.0 and 4.0.

NRC POSITION 12c. The PORVs perform their function before being exposed to the harsh environment, and the adequacy of the time margin provided is justified; subsequent failure of the PORVs as a result of the harsh environment will not degrade other safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifi-cally prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed under harsh environment I

conditions).

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OPPD RESPONSE I The PORVs function is discussed in response to 12a, see Sections 1.0, 2.0, 3.0 and 4.0.

NRC POSITION

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! 12d. The safety function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-failure criterion.

OPPD RESPONSE See response to 12a, Sections 5.0 and 6.0.

CONCLUSION OPPD believes that the previously discussed configuration is adequate to insure safe operation in all plant operating modes requiring PORV operation and that adequate control and indication has been provided to mitigate potential accident failure modes of the PORVs.

a O ATTACHMENT 1 I

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September 14,1982 G-7610-190 To: C. R. Harvey D. Johnson CV-45 7A44 cc: R. W. Blohm R. D. Broad 7A.21 E. 3. Corrie 6K-39 D. P. Konichek 7A-20 R. C. Lundquist 9A-02

3. F. Frestl 7A 36 -

M. 3. Synge 7A-21

3. L. Tocher 7A.21

3. C. Turley 9C-02 R. Vontoble 7A.23 6K-39

Subject:

Certification: Force V2 .

The FORCE program is certilled to perform as described in attachment 1.

Technical Requirements for Class B, Regulated, described in document 40356.01, have been med as evidenced in attachment 2. -

Conditional Certification, Class B and Category Reguinted is granted EECCL, Nucilbe Vendor, for development and main FORCE will be Installed on the EK5 Mainstream and V5P Services,

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, - ETA Qua,llty Assurance Mm F. A. Hanna Engineering and Scientific Services Attachments

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Attachment 1 1.

MAINSTREAM-EKS, 2032, July 1982 FORCE, Reference Manual and Access Guide,10208-2.

Test Report, FORCE, G 7623-046, August 27,1982

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. w , . .. .. en November 1.1982 - Revision C-7623-046R To B. Block F. Hanna B. Mukher11 S. Pruitt C. Wolfe

Subject:

Certification (QA Section 2.3) FORCE Version 2 Quality A Refer 2nce: Memo G-7623-028 June 28,1962 (QA Section 2.1.2) Test Plan for RELAP 5/1 and FOR Test Procedure Execution Results (OA Section 2.3.2)

The test system as planned.cases set forth in the referenced Test Plan were run on th on tape (Attachment All flies used A). The testtocase create runs,and test this including version input have are and output, beenbo in and"FORCE are a part of Quality Assurance Standard Test Case Set and Hand Calculat this certification.

. Test Analysis (QA Section 2.3.3) 1 The 2.1.2), QA results7, QA-9, of andthe QA-10three test as are presented cases follows: cited in the standard Te 1.

Model) case was run to provide test output from c and to demonstrate ability to handle safety release valve tests. Cursor examination of output Indicates that the code handles this case successfu 2.

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results matched the Combustion engineering and E attached shows plots the an ear,ller valves being comparison donethe same as the "BCS" curve. Attachment B replicates these results for FORCE certification.for verification and Attachment C 3.

Case QA-10 la a simplified model of a pipe and it was run to provide a manageable hand calculational case Intended to reinforce that FORCE ,

performs its calculations properly.

pipe based on fluid and gas conditions. FORCE calculates forces in a hydraulle The calculations are made from density, velocity, pressure, and time parameters and the geometry of segments under consideration. A run of this case on the RELAP 3/1 code was made and the output used as input to FORCE.

comparison with the run are shown in Attachment O The calculations and

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both input and output, are bound as part of the certifi Test Deficiencies (QA Section 2.3.4)

No deficiencies were found in the testing of FORCE.

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(see QA Section 3.8).Any defielencies discovered in future will be given in

'$'k$rs ..'ebcl D. P. Konichek Attachments:

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l Af y pfprp Ntober 8, 1984

!4-DAJ-103 Fctr Cy 11 To: J. C. Jervert i"

Subject:

T PIPE, Regulated BCS Version 2.1--Hational Certification Class A and Category Reference-Memo G-7430-MAG-024, Dated November 30, 1979, M. A. Grece to C. S. Bartholomew, et al., Subject T PIPE Release Certification T-PIPE, BCS Version 2.1, was previously certified with no qusll'ications per the reference memo.

Subsequently, the category of "Regulate,4" was used to define a set of certified products that were to be used by the nuclear industry customer set and had Procedures completed 40356.01 the certification process as defined by Quality Assurance series.

T-PIPE, BCS Version 2.1, has been certified by this process lated." and has met all requirements for certification to the category of "Regu-T PIPE is a program which perfonns stress analysis of piping systems. It ;rovides static analysis, dynamic analysis, NRC Regulatory Guide methods, ASME Class I thermal transient analysis, and1.92 mode com:ination stress classification ac-cording to ASME and ANSI B 31.1. Boiler and Pressure Yessel Code Section III, Class 1,'2, and 3 1

This letterthe regarding is category issued for the product.

of this purpose of upgrading the Quality Assurance records i

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i O. A. Johnson, Manager

, Headquarters Quelity Assurance 7C-36, 763 5122 DAJ:syv

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November 30, 1979 G-7430-MAG-024 f'

To:

S. Bor Geivmew R. J. Flynn K. D. Johansen S. L. S. Jacob D. H. La.ngdah)y cc:

E. J. Corrie M. J. Syngs J. L. Tocher F. L. Hisa .

Subject:

T-PIPE Releas! Certificati Data Package is attached.The status of each of the items ide Applications Division including configuration man.agement. is responsible for T-PIPE Product S It is theberecomendation T-PIPE certified as a BCScf theAT-PIPE Class Certification product with. C full cartificat M. A. Groce Y 14-

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T-PIPE Certification Committee Chairman Attachment

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