NLS8800476, Application for Amend to License DPR-46,deleting 1,000 Psig Pressure Permissive from MSIV Closure Scram Trip Setting & Making Miscellaneous Editorial Corrections.Fee Paid
ML20205P502 | |
Person / Time | |
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Site: | Cooper |
Issue date: | 11/01/1988 |
From: | Kuncl L NEBRASKA PUBLIC POWER DISTRICT |
To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
Shared Package | |
ML20205P507 | List: |
References | |
NLS8800476, NUDOCS 8811080184 | |
Download: ML20205P502 (11) | |
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Nebraska Public Power District Nhk S$fhst"""
NLS8800476 November 1, 1988 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:
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Subject:
Proposed Change No. 51 to the Cooper Nuclear Station Technical Specifications NRC Docket No. 50-298, DPR-46 In accordance with the applicable provisions specified in 10CFR50, the Nebraska Public Power District requests that the Cooper Nuclear Station Technical Specifications be revised as indicated in Attachment 1. The purpose of this change is to delete the 1000 psig pressure permissive from the MSIV closure scram trip setting and to make several miscellaneous editorial corrections.
Attachment 1 contains a description of the proposed changes and the results of the evaluation of the proposed changes with s
respect to the requirements of 10CFR50.92. The applicable revised Technical Specification pages are also attached.
By copy of this letter and the attachment, the appropriate State of Nebraska Official is being notified in accordance with 10CFR50.91(b).
This proposed change incorporates all amendments to the Cooper Nuclear Station Facility Operating License through Amendment 125 issued August 8, 1988.
This change has been reviewed by the necessary Safety Review Committees and payment of $159 is submitted in accordance with 10CFR170.12.
In addition to the signed original, 37 copies are also submitted for your use. Copies to the NRC Region IV Office and the CNS Resident Inspector are also being sent in accordance with ol 10CFR50.4(b)(2). fo
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. NLS8800476 Page 2 ,
November 1, 1988 ,
e Should you have any questions or require' additional information, please contact this office.
Since ly, M M L. G. Kunci
- Nuclear Power Group Manager '
LGK/mtb:dar3/4 i Attachment cc: H. R. Borchert Department of Health ;
State of Nebraska NRC Regional Office Region IV ;
Arlington, TX :
i NRC Resident Inspector Office Cooper Nuclear Station 1
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. NLS8800476 Page 3 November 1, 1988 STATE OF NEBRASKA)
)ss PLATTE COUNTY )
L. G. Kuncl, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this request on behalf of Nebraska Public Power District; and that the statements contained herein are true to the best of his knowledge and belief.
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L. G. Kunc{l Subscribed in my presence and sworn to before tr.a on this _lllb 1henJr0
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- Attcch tit to-NLS88004/6 Revised Technical Specifications for Change No. 51 Revised Pages: .9 70 75 108 168 Proposed Change No. 51 to.the Cooper Nuclear Station Technical Specifications involves four separate miscellaneous revisions. The four proposed revisions are:
- 2) T.S.4.4.A.2.c, page 108, revised testing requirements for Standby Liquid Control System explosite valves. *
- 3) Table 3.7.1, page 168, correct containment isolation valve table, and-
- 4) Table 4.2.B. pages 70 and 75, delete relay calibration requirements.
The first of the four changes affects Limiting Safety . System Setting ;
(LSSS) 2.1. A.5 on page 9 of the Cooper Nuclear Station (CNS) Technical Specifications. This change would delete the pressure permissive from the c MSIV Closure Scram Trip Setting. The current Technical Specification LSSS !
requires a scram trip signal prior to 10% MSIV closure, when above 1000 psig reactor pressure, in 3 out of 4 main steam lines. This change would delete ,
4 "when above 1000 psig reactor pressure," and add ",and the reactor mode switch is in the "Run" position."
The putpose for this pressure permissive in the MSIV closure scram trip I setting was, as stated in GE Report NEDC-20546, "Cooper Unit 1 Transient i Analysis Design Report," to ensure that the MSIVs were more than 90% open when j pressure exceeds 600 to 1000 psig in the startup or hot standby mode.
General Electric (GE) reviewed this MSIV closura scram pressure permissive and .
determined that the pressure permissive was not necessary. This evaluation I was documented in Design Information Memo (DIM) No. 131 dated September 12, l 1974. This DIM states that the pressure permissive was installed as a result j of experience at KRB in Germany, where during startup, operations had !
difficulty controlling reactor power above 600 psi without pressure control. ,
! A test was conducted during the Browns Ferry Unit I startup (Document 22A2510) l which showed that plants similar to Browns Ferry, such as Cooper, did not have i the problems exhibited at KRB. GE concluded that the MSIV closuro scram is ,
only roquired when the reactor mode switch is in the "Run" position, and the pressuro permissive is not required.
Additionally, the current high reactor vessel pressure scram setpoint is f
< 1045 psig. Maintaining an MSIV closure scram permissive of 1000 psig does :
not significantly increase safety in view of the existing, essential high ,
pressure scram.
Attachm:nt to NLS8800476 Page 2 General Electric subsequently recommended the use of 6he MSIV closure scram bypass pressure switches in the Low Low Set Relief logic. This design for Low Low Set was documented in GE Report NEDE-22223, dated September 1982. The District submitted NEDE-22223 to the NRC, along with e description of the LLS design change, in a [[letter::A820003, Application for Amend to License DPR-46 Changing Tech Specs, Safety Relief Valve (S/Rv) Low-Low Set (LLS) Sys & Lower MSIV Water Level Trip. Encl Withheld (Ref 10CFR2.790)|December 17, 1982 letter]], fron J. M. Pilant to D. B. Vassallo. The Dictrict also submitted on the docket NEDE-22107, "Low-Low Set Relief Logic System and Lower MSIV Water Level Trip for Cooper Nuclear Station, Unit 1," in a letter 1 ted February 15, 1983, from J. M. Pilant to D. B. Vassallo.
The NRC approved the GE generic Low-Low Set logie design in a letter f t'om D. B. Vassallo to H.C. Pf ffarlen, dated April 26, 1983. The NRC air.o approved the Cooper spet a . i 'v-Low Set design and Technical Specifiention changes, in License Amendment 'a. 83, dated May 4, 19 P.3.
The District clearly descrit.c the removal of the 1000 psig pressure permissive from the MSIV closure scram trip setting and the use of these pressura switches in the Low-Low Set modification in the December 17, 1382, letter. The NRC was rognizant of this change and provided prior approval of the modification in 'he April 76, 1983, letter. However, the appropriate change to remove the 1000 psig permissive from LSSS 2.1.A.5 was inadvertently omitted from the technical specification changes that were cubmitted in connection with the Low-Low Set modification.
This change is intended to remove the 1000 psig pressure permissive from the MSIV closure scram, which corrects the oversight in Amendment 83 (Low-Low Set). This changJ does not involve any change to any hardware or operating procedure.
The second proposed change would revise Section 4.4. A.2.c on page 108 of the CNS Technical Specifications. This section requires the actuation of the explosive charge in one of the two loops of Standby Liquid Control once each operating cycle. The replacement explosive valve is currently required to come freu the same manufacturer's batch as the valve just test fired. This proposed change would revise Specification 4.4.A.2.c to allow the replacement explosive valve to ecme from a previously tested manufacturer's batch.
This change is intended to reduce the number of explosive vsives from each batch that must be test fired, without reducing the reliability of the replacement valves. Test firing one valve from each batch is adequate to de=enstrate reliability of the charges in that batch. As the specification currently reads, replacement valves must come from the same batch as the valve that was just test fired. The proposed change would allow installation of a replacement valve as long as one valve from that batch has previously beon tesc fired. This will reduce the number of explosive valves test fired.
The thirL proposed change revises Table 3.7.1 on page 168 of the CNS Technical Specifications. The Component Identification Code for one of the Drywell Floor Drain Isolation Valves is corrected from "RW-AC-83" to "RW-AO-83."
Also, the current Technical Specifications show the Drywell Floor Drain Isolation Valves to be ene inboard and one outboard valve. The configuration at CNS actually consists of tuo outboard valves and no inboard valves. The same configuratien exists for the Drywell Equipment Drain Isolation Valves ,
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- NLS8800476 Pags 3 i RW-AO-94 and RW-A3-95 and the RHR Discharge to Radwaste Isolation Valves, RHR-MO-57 and RHR MO-67. Therefore, a correction to Table 3.7.1 is required i to show two outboard valves and no inboard valves for RW-AO-94 and 95 and RHR-MO-57.and 67, The configuration of the drywell floor drain and equipment drain isolation ,
valves is in accordance with the requirements for Class B isolation valves as ,
described in Chapter V, Section 2.3.5.1, of the Updated Safety Analysis Report ,
(USAR). Class B valves are defined in the USAR as those valves on process b lines that do not directly communicate with the reactor vessel, but penetrate the primary containment and communicate with the primary containment free 4
space. Thus, the drywell floor and equipment drain isolation valves are Class B valves. In accordance with Section V-2.3.5.1 of the USAR, Class B valves are to be in series - and both outside the primary containment. [
Therefore, the exf. sting configuration is in accordance with the CNS design and i licensing basis.
The RHR discharge to radwaste isolation valves, RHR-MO-57 and 67, are located on a 4 inch branch line connected to the 20 inch cross-tie line between the two loops of RHR. Although the 4 inch branch line is connected directly to the reactor vesse:. by way of the 00 inch cross-tie and the RHR (LPCI) piping, ;
' in all cases there are two other isolation valves in series between MO-57 and l j 67 and the reactor vessel. MO-57 and 67 provide for isolation of the RHR [
piping as it leaves the reactor building and unters the radwaste building. f
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Due to the location of this branch connection, the only physical configuration !
! possible is two isolation valves ootside .:ontainment. This change, therefore,
- l revises the Technical Specifications to reflect the actual plant !
The current plant configuration is appropriate, since MO-57 configuration.
and 67 are the second set of isolation valves in series, used to isolate a (
4 inch branch connection off of the cross-cie line.
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, This change does not involve any hardware or procedural changes. The !
Technical Specifications contain editorial errors that should be corrected to [
j match the actual configuration of the plant.
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j The final proposed change revises Table 4.2.B on pages 70 and 75 of the CNS i
, Technical Specifications. The purpose of this change is to delete the j
! calibration requirements for relays 27X1/1T(1G), 27X2/1T(10), 27X3/1A(1B), .
- 10A-K79A(B) and 10A-KSOA(B). (
I j These relays are auxiliary devices. They operate as a result of signals [
(contact closures) provided by another primary device, such as a level f
! indicating switch. These relays are either on or off. They do not provide a j i variable output, and therefore, they cannot be calibrated. :
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- In the case of relays 10A-K79A(B) and 10A-K80A(B), the primary device, ;
i NBI-LIS-72(A.B.C.D) is calibrated on a three month frequency. The primary ,
l device provides a signal to the ralays in question in response to changes in !
j the process variable. The 10A-K79A(B) and 10A-K80A(B) relays are functionally ,
j tested during calibration of the primary device. ;
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Attcchm3nt to NLS8800476 Page 4 The Emergency Bus and Auxiliary Bus undervoltage relays 27X1/1F(1G),
27X2/1F(1G) and 27X3/1A(1B) are also either on or of f. They cannot be calibrated since they do not provide a variable output. These relays are
- functionally tested as required by Technical Specification 4.2.B, but calibration is not appitcable. ,
Evaluation of this Amendment with Respect to 10CFR50.92~
, A. The enclosed Technical Specification change is judged to involve no significant hazards based on the following:
- 1. Does the proposed license amendment involve a significart increase in the probability or consequences of an accident previously evaluated?
Evaluations
- a. The first proposed change deletes the 1000 psig pressure The permissive from the MSIV Closure Scram Trip Setting.
original design of the Cooper Nuclear Station (CNS) incorporated a scram trip setting to initiate a scram signal,if the reactor pressure exceeded a setpoint (coe=only 600 psig) with the MSIVs closed and the reactor mode switch in "Startup."
The design feature was the result of experience at a Bk'R in
, Germany (KRB) . At KRB, operators had difficulty controlling reactor power during startup above about 600 psig without i pressure control.
l In approximately 1974, Bk'Rs in the U.S. recognized the
! advantages of being able to heat up to rated pressure with the MSIVs closed. This provides the capability to heat the vessel and internals without running the feedpumps and allows heatup j in parallel with work on the turbine. General Electric
! conducted tests at Browns Ferry Unit 1 (document 22A2510) which j demonstrated that heatup to rated pressure in s "bottled-up" i condition, is within the capability of the plant. GE issued i
Design Issue Memo (DIM) No. 131, dated September 12, 1974,
! which documented the technical justification for removing the i MSIV closure scram pressure permissive and identified the Brown l Ferry test results as applicable to CNS. DIM No. 131 l recommended either removing the pressure permissive hardware or
- increasing the pressure setpoint to the high ' pressure scram point.
i The CNS Technical Specifications were changed accordingly from I
the original 600 psig setpoint to 1000 psig, which is near the 1045 psig high pressure scram setpoint. During the Low-Low Set modifications, these pressure switches were used and deleted from the logic for the MSIV Scram Closure Trip. However, the associated Technical Specification change to delete the 1000 psig pressure permissive was not made.
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Attrchment to NLS8800476 Page 5 As documented in GE Design Information Memo (DIM) No. 131, the original purpose for the pressure setpoint in the MSIV Scram Closure logic no longer exists. The testing conducted at Browns Ferry and DIM No. 131 document that the pressure setpoint can be deleted. Also, the analyses conducted for the Low-Low Set Modification were submitted to the NRC and approved.
This change, therefore, only corrects the editorial error in the CNS Technical Specifications. No hardware changes are involved. Since the Change revises TS 2.1.A.5 to match the existing plant conf *guration, this change does not involve an increase in the probability or consequences of an accident previously evaluated.
- b. The second propos2d change revises the testing requirements for the Standby Liquid Control (SLC) System explosive valves. This proposed change does not affect the intent of Specification 4.4.A.2.c which is to ensure that each explosive valve charge installed comes from a batch where at least one charge has been test fired. TS 4.4.A.2.c currently requires that each valve installed must come from the same batch as the valve just test fired. This can double the number of charges that must be test fired, without improving the reliability.
Testing one charge from each batch is adequate to verify the batch. The Standard Technical Specifications already contain the provision which allows the new valve to come from a previously tested batch. This change does not affect the frequency of testing the firing circuit or the SLC system.
Since this change clarifies the Technical Specification, and agrees with the Standard Technical Specificationu, but does not affect test frequency of the SLC System or the reliability of the charges from a manufacturer's batch, no increase in the probability or consequences of an accident is involved.
- c. The third proposed change cakes editorial corrections to Table 3.7.1, which contains a list of Containment Itolation Valves. The revisions include correcting a typographical errer in the ID Number of a Drywell Floor Drain Isolation Valve.
AW-AO-83, and correcting the configuration of the Drywell Floor D ain. Dryvell Equipment Drain and RHR Discharge to Radwaste Isolation Valves from one inboard and one outboard valve to two outboard valves.
The change to correct the configuration of tho three sets of isolation valves revises the Technical Specifications tu match the actual configuration at CNS. The Drywell Floor Drain and Equipment Draia Isolation Valves are Class B isolation valves as defined in the USAR. The two outboard valve configuration is correct for Class B isolation valves in accordance with the Updated Safety Analysis Report (USAR). For the RHR Discharge to Radwaste isolation valven, the two outbeard valve configuration is also appropriate, due to the physical location of the branch connection being isolated and the fact that these valves are the sacond set of isolation valves in series.
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Atttchm:nt to
. NLS8800476 Pcgo 6 Based on the above,' correcting the typographical error and ,
correcting the valve configuration in Table 3.7.1 to match the actual plant configuration do not increase the probability or
, consequences of an accident.
- d. The final proposed change would delete the calibration requirements for the reactor low water level relays, 10A-K79A(B) and 10A-80A(B), and unde rvoirage relays, 27X1/1F(1G), 27X2/1F(1G) and 27X3/1A(IB). This change does not affect the current hardware or surveillance procedures. It is not possible to calibrate these relays since they are "on" or "off" type relays. There is no variable output. It is only possible to functional *y test these relays to ensure that they .
pass on the signal when received from the transducer. These inte rmediate relays are already functionally tested in accordance with the current Technical Specifications. The functional test frequency is adequate and is not being changed as a part of this revision.
This change does not affect the reliability of the relays in question, and therefore, does not increase the probability or consequences of an accident.
- 2. Does the proposed license amendment create the possibility for a new or different kind of accident from any accident previously evaluated?
Evaluations
- a. The modification which deleted the 1000 psig pressure t
permissive from the MSIV Scram Closure logic was completed '.n e1983. Since the current setpoint for the MSIV Closure Scram pressure permissive was 1000 psig and the high pressure scram setpoint is 1045 psig, it is clear that the high pressure scram will fulfill the function previously performed by the MSIV pressure setpoint. Further, tests and analyses documented by General Electric have shown that the original purpose of the
- pressure setpoint is no longer valid. The MSIV Closure Scram l is only applicable in the "Run" mode.
The original design and licensing basis for the 1000 psig
- pressure setpoint (originally set at 600 psig) was to prevent heatup in a "bottled-up" condition. It has been shown chrough testing that the Cooper Nuclear Station design will accommodate a "bottled-up" hestup. The modification to use the 1000 psig pressure switches from the MSIV Scram Closure logic in the Low-Low Set modification was analyzed and submitted to the NMC for approval. This modification received prior Commission approval. This change revises the Limiting Safety System Setting to agree with the current plant configuration. Since the change to the Technical Specificati m revises LSSS 2.1.A.5 to agraa with tbv actual plant configera, ion and the existing plant configurteion has been shown by test to be safe, this change does nat create any new or different kind of accident.
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, Attcchm:nt to NLS8800476 ^
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- b. This change clarifies the wordin6 to allow a charge from a previously tested batch to be installed instead of requiring a charge from the samt batch just tested to be installed. The replacement charge will be from a tested batch and the test frequency has not been changed. No reduction in the reliability of the replacement charges will result as demonstrated by the fact that the Standard Technical Specifications already contain this provision.
Since the type of replacement charge has not been changed and no changes have been made to the Standby Liquid Control System hardware or testing, there is no new or different kind of accident created.
- c. The changes to Table 3.7.1 are editorial in nature. The changes include correcting an ID Number and correcting the configuration of three sets of isolation valves. These are not hardware changes and do not affece the operation of the valves in any way. The existing plant configuration for the Drywell Floor and Equipment Drain isolation valves is in accordance with the PSAR. The RHR Discharge to Radwaste isolation valves are configured appropriately, given the physical location of the branch line being isolated and the fact that there are two other valves in series closer to the reactor vessel. All of the proposed changes are, theref?re, editorial in nature and de not create any new or different hind of accident.
- d. The final proposed change, to delete calibration requirements for "on" or "off" intermediate relays, also involves no hardware changes. The relays in question cannot be calibrated; they can only be functionally tested. The functional testing is already conducted in accordance with the Technical Specifications. Since this change only deletes a calieration, which has no meaning for these relays, no i.ew or different kind of accident is created.
- 3. Does the proposed license amend =ent involve a significant reduction in a margin of safety?
- a. The 1000 psig pressure setpoint that is currently included as a part of the MSIV Closure Scram was physically removed in 1983.
The basis for eliminating the pressure permissive was testing conducted at Browns Ferry. This testing showed that the original instability problem during startup is not a problem in BWRs like Browns Ferry and Cooper. Further, the high pressure scram setpoint is already set at 1045 psig. Therefore, correcting the Technical Specificatious to delete the 1000 psig pressure permissive does not reduce the margin of safety,
- b. The change to Technical Specification 4.4. A.2.c clarifies ':he testing requirements for the explosive valves in the Standby Liquid Control (SLC) System. This change does not affect the intent of the surveillance requirement. The result of the proposed change is that replacement charges must be from a previously tested bateh, instead of from the same batch as the charge just test fired.
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. l Attcchment to
. NLS8800476 Page 8 This does not reduce the testing frequency of the SLC system or reduce the reliability of the explosive va.ves. If the charge installed in the system at the time of testing is the last charge in a batch, then two charges must be test fired and a third one installed as a replacement. This is clearly not the intent of the surveillance requirement. It is intended that the replacement charge come from a tested batch, which is the result of thir proposed revision. Also, the Standard Techaical Specifications have been approved with the provision that the new charge must come from a previously tested batch.
Based on the above, this change does not reduce any margin of safety.
- c. The editorial changes to Table 3.7.1 correct an ID Number and correct the configuration of three sets of isolation valves.
The ID Number is purely editorial. The isolation valves will be changed to match the actual configuration in the plant. The existing plant configuration for the Drywell Floor and Equipment Drain isolation valves agrees with the requirements for isolation valves in systems which communicate with the drywell atmosphere as stated in the Updated Safety Analysis Report. The configuration of the RER Discharge to Radwaste isolation valves is appropriate for their function. Therefore, this proposed change does not reduce the margin of safety,
- d. The final change deletes the calibration requirements for several "on" or "off" type relays. Calibration of these relays has no meaning. They are functionally tested which verifies the safety function, on an approved Technical Specification frequency. Deletion of this meaningless requirement,
- therefore, does not reduce the margin of safety.
B. Additional basis for proposed no significant hazards consideration determination:
The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48CTR14870). The examples include: "(1) A purely administrative change...and (iv) A relief granted upon demonstration of acceptable operation from an operating restriction...." Proposed change (a) fits under example (iv), since the original instability problem that was the reason for the 1000 psig pressure permissive, was subsequently shown by test not to be a concern at Cooper. Also, proposed change (a) fits under example (1), since this change revises the Technical Specifications to match hardware changes made in accordance with a previously NRC approved modification. Proposed i changes (b), (c) and (d) are all considered to fit under example (1).
l Change (b) clarifies the testing requirements, without affecting the I intended purpose of the test, to reduce needless test firing of explosive valve charges. Change (c) corrects an ID Number and valve configuration l to match the actual, correct plant configuration. Change (d) deletes
! meaningless calibration requirements for "on" or "off" type relays.
l These are all purely administrative editorial corrections.
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