ML20127K328

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SER Accepting Amend 9/change 18 to DPR-22 Correcting Typos
ML20127K328
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 04/10/1975
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20127K314 List:
References
NUDOCS 9211200374
Download: ML20127K328 (10)


Text

_ _ _ _ _ _

UNITED ST ATES NUCLEAR REGULATORY COMMISSION W ASHING 1oN. o. C. 20$$$

SAFETY EVAI.UATION BY Tile OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 9 TO FACILITY ODERATING LICENSE NO. DPR-22 (CilANGE NO.18 TO Tile TECilNICAL SPECIFICATIONS)

NORTl!ERN STATES P0h'ER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 INTRODUCTION Sy letter dated November 15, 1974, Northern States Power Company (NSP) proposed changes to the Technical Specifications of Provisional Operating License No. DPR-22 for the Monticello Nuclear Generating Plant. Item 16 of the licensee's proposed change, which concerns the reactor vessel material specimen withdrawal program, was approved on February 3,1975. The staff finds that NSP'c proposed change number 9.b (Item 21 of this Safety Evaluation) is not acceptable. The basis for our disapproval of the above proposed change is that it would permi reactor operation up to approximately fifteen percent of full rat ed power, instead of the presently approved one percent power rest rict ion, without requiring the isolation valves listed in Table 3.7.1 of the Technical Specifications to be operable. The corresponding increase in the reactor coolant system stored energy could make the radiological consequences of a le s s-o f- coo l an t accident coincident with inoperable isolation valves unacceptable. The remainder of the proposed changes, many of which simply involve typographical corrections, are the subject of this evaluation.

EVALUATIO:s

1. The changes to the Table of Contents, List of Figures and List of Tables (7 pages, i.e., pages i through vii) are corrections of typo-graphical errors and updating to add previously approved changes to the Technical Specifications.
2. Page 38 - The change on this page is a typographical correction.
3. Page 39 - The statement, " Discharge of excessive amounts of radio-activity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation to the main condenser.. ." is no p.or eog e

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longer applicable. Prior to the installation of the Augmented Off-Gas System, which consists of two redundant catalytic recombiner trains

) and a compressed gas storage system, which provides a 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> delay time, the air ejector off-gas monitors would terminate gaseous j release to the environsby automatic isolation of the condenser off-gas system. Subsequent to the installation of the Augmented Off-Gas System, isolation of the condenser would not terminate release

' of gases to the site environs because the contents of the compressed gas storage tanks, which are located downstream of the main condenser l off-gas isolation valve, would continue to discharge to the site environs if it were in the purge mode. Termination of gas storage tanks discharge will occur automatically upon receipt of a closure signal from the stack radiation monitor which would cause closure

of the stored gas isolation valve. The net result is that the air-4 ejector off-gas monitors would retain their original isolation i roles only in the event the Augmented Off-Gas System were bypassed.

Since the installation of the Augmented Off-Gas System has been previously approved, we conclude that the licensce's proposal is acceptable. A description of air ejector monitors functions is provided in page 68 of the currently approved Technical Specifications.

4. Page 40 - This change was made to further clarify that the high reactor pressure,- high drywell pressure, and reactor low water level and scram discharge icvel trip functions may be bypassed when in the " Refuel" mode with the reactor suberitical and the reactor coolant temperature less than 212*F as allowed by Note 3 of Table 3.1.1 (Page 31). For clarity, NSP's proposed wording has been slightly modified.
5. Pages 43 and 4S - The changes on these pages climinates an unnecessary reference since the failure rate statistics are as listed in the Bases immediately preceeding the reference.

! 6. Page S9 - The changes in this page are a typographical correcti= and a change to allow bypass of the source range monitor (SRM) upscale block when the associated intermediate range ronitor (IRM) range switches are above Position 6 instead of above Position 7. The SD1 Upscale rod block is required only when insufficient range overlap exists on the IRM before the SRM's reach the upper range 4

limit. At least one decade, but preferably two decades, of overlap is sufficient to assure a definite overlap and transition from monitoring of SRMs to monitoring _IRMs. Since Position 6 corresponds-to a three decade overlap, the change is considered acceptable. In

- addition, operating experience has revealed that the original bypass above the Position 7 is overly restrictive and interferes with normal startup in a manner that was not intended.

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7. Page 62 - The change on this page is a typographical correction.
8. Page 63 - Two additional sentences are added to Note 2 (for Table 4.2.1) to define SRM and IRM calibrations to avoid misunderstanding.
9. Page 78 - An unnecessary date (May 1,1974) is removed from the limiting conditions of operation. The surveillance requirements are changed to require rod position verification prior to insertion as well as withdrawal of each rod group to correct an omission error.
10. Page 85A - The change on this page is a typographical correction.
11. Page 89 - The primer assembly contains two charges that cannot be separated for test firing. The change reflects the reality of the single primer assembly containing two charges rather than independent charges as conceived when the Technical Specification was written. The change eliminates the misleading impression that the explosive charges can or must be tested individually and is accepteble.
12. Page 94 - Dilution from water in the cooldown circuit was neluded in the original determination of standby liquid control system requirements. The change to the Technical Specification bases justifiably acknowledges this to avoid uncertainty about basic design information,
13. Page 104 - Corrects numerical errors in converting absolute pressure to gauge pressure.
14. Page 108 - The change on this page is a typographical correction.
15. Pages 116-118 - Our evaluation of NSP's proposal for a surveillance requirement change is as follows:

(1) Sampling of reactor coolant for radioiodines of I-131 through I-135 instead of sampling for gross beta activity is approved based on the current Standard Technical Specification format.

(2) We instituted a limiting condition of 5 microcuries I-131 dose equivalent per gram of water to replace the 20 microcuries of total iodine per cubic centimeter of water. This change was made to make the radiological consequences of a steam line break outside of containment acceptable and well within the limits of 10 CFR 100.

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Additionally, we instituted surveillance requirements for steady state radiciodine concentrations prior to operation to be consistent with that of similar boiling water reactor installa-tions. The change reflects a means for surveillance of possible transient radioiodine spike conditions during startup and shutdowns.

Our other changes reduced the reactor coolant water chemistry ,

conductivity and chloride limiting conditions in accordance with '

Regulatory Guide 1.56 (June 1973). The new limits reduce the probability of corrosion induced failures in reactor coolant boundaries in boiling water reactors. A reactor shutdown, if reactor coolant water purity limits are exceeded, assures curtail-ment of reactor operations until suitable corrective measures are taken. All of the above mentioned changes were discussed with NSP (phone con of January 28, 1975) and found to be acceptabic to NSP.

16. Page 118 - The change improves clarity by adding "during power operation". NSP inadvertently left the item out of their original submittal.
17. Pages 132 and 133 - Our change provides the basis for the Suci/ml of water I-131 dose equivalent limit that replaced the 20pCi/ml total iodine limits in specification 3.6.C.1, i.e., the steamline break accident outside of containment. It is in the standard Technical Specification format and has been reviewed and found acceptable by NSP.
18. Page 147A -The change utill:es tne Monticello Technical Specification definition of "run mode" to simplify the specification operating limit without compromising the basic requirements for containment inerting as stated in 3.7.A.5.b.
19. Page 148 - The change permits filter tests 30 days prior to Therefueling existing so that related activities can be scheduled efficiently.

requirement that filter performance tests be conducted during the refueling outage is without basis and unintentionally restrictive.

20. Page 150 - No action was taken on this item because the typographical error did not appear on page 150 as revised with Change No.12 issued by the Commission on November 15, 1973.
21. This proposed change was disapproved for reasons stated in the introduction to this evaluation.

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22. Page 152 - The change in 3.7.D.2 to use "run mode" does not alter the intent of the limiting conditions of operation with respect to continued reactor operation with inoperable isolation valves.

The change in 3.7.D.3 corrects an oversight by properly referring to specification 3.7.D.1 and 3.7 D.2 instead of 3.7.0 only.

23. Page 170 - The change on this page is a typographical correction in specification 3.8.A.9. The changes to specification 4.8. A.4 require accilerated surveillance at the point of increased iodine or particulate releases only, and are similar to surveillance require-ments for other BhRs with separate stack and vent releases.
24. Page 172 - The change eliminates specification 4.8.C.2e because The it is repetitious, i.e., specification 4.8.C.1 is adequate.

NSP proposed page change mistakenly ocleted 4.8.C.2.d although item 20 of Exhibit A of NSP's request dated November 15, 1974, correctly identified Specification 4.8.C.2.e. This has been confirmed by discussion with NSP representatives (phone con January 29, 1975).

25. Page 181 - The change restores the phrase "or the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" which was unintentional 1 deleted from the specification previously. The correction is required to specify the action that must be taken in the event that sources of electric power are unavailable.
26. Page 183 - The change requires a shutdown and cooldown to 212*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one 125V battery system or the 250V battery system l is unavailable. The proposed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown requirement is consistent with the safety importance of this system and justifiably more restrictive on plant opetation than the 7 day period currently specified.

CONCLUSION We have concluded, based on the reasons discussed above, that because the changes do not involve a significant increase in the probability or involve a consequences of accidents previously considered and do not significant decrease in a safety margin, the changes do not involve a significant hazards consideration. We also conclude that there is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: April 10, 1975

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l UNITED STATES NUCLEAR REGULATORY COMMISSION

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I SAFETY EVALUATION BY Tile OFFICE OF NUCLEAR REACTOR REGULATION 3

] l 1 SUPPORTING AMENDMEffr NO. 9 TO FACILITY OPERATING LICENSE NO. DPR-22 l

l (CHANGE NO.18 TO illE TECHNICAL SPECIFICATIONS)

NOR11 FERN STATES POWER COMPANY t

MONTICELLO NUCLEAR GENERATING PLANT

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DOCKET NO. 50-263 i

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! INTRODUCTION l

'By letter dated November 15, 1974, Northern States Power Company (NSP)

- proposed changes to the Technical Specifications of Provisional Operating ,

i License No. DPR-22 for the Monticello Nuclear Generating Plant. Item 16 of the licensco's proposed change, which concerns the-reactor vessel material specimen withdrawal program, was approved on February 3, 197S. The staff i finds that NSP's proposed change number 9.b (Item 21 of this Safety hvaluation)

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is not acceptable. The basis for our disapproval of the above proposed change j

is that it would permit reactor operation up to approximately fifteen percent 4

of full rated power, instead of the precently approved one percent power

" restriction, without requiring the isolation valves listed in Tabic 3.7.1 of the Technical Specifications to be operable. The corresponding increase in

" the reactor coolant system stored. energy could make the radiological consequences 4-of a loss-of-coolant accident coincident with inoperable isolation valves unacceptable. The remainder of the proposed changes, many of which simply

involve typographical corrections, are the subject of this evaluation.

1

[ EVALUATION

1. The changes to the Tabic of Contents, List of Eigures and List of Tables (7 pages, i.e., pages i through vii) are corrections of typo-

- graphical errors and updating to add previously approved changes to the Technical Specifications.

- 2. Page 38 - The change on this page is a typographical correction..

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3. Page 39 - The statement, " Discharge of excessive amounts of radio-activity to the site environs is prevented by the air ejector off-gas- .

monitors which cause an isolation to the main condenser..." is no .

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s longer applicable. Prior to the installation of the Augmented off-Gas

! System, which consists of two redundant catalytic recombiner trains and a compressed gas storage system, which provides a 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> delay time, the air ejector off-gas monitors would terminate gaseous

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release to the envirornby motonatic isolation of the condenser off-gas system. Subsequent to the installation of the Augmented Off.

~ Gas System, isolation of the condenser would not terminate release of gases to the site environs because the contents of the compressed gas storage tanks, which are located downstream of the main condenser off-gas isolation valve, would continue to discharge to the site environs if it were in the purge mode. Termination of gas storage

' tanks discharge will occur autmaatically upon receipt of a closure signal from the stack radip*, ion monitor which would cause closure of the stored gas isolation valvee The not result is that the air-ejector off-gas moniters would retain their original isolation

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roles only in the event the Augmented Off-Gas System were bypassed, since the installation of the Augmented Off-Gas System has been previously approved, we conclude that the licensee's proposal is acceptable. A description of air ejector monitors functions is provided in page 68 of the currently approved Technical Specifications.

4. Page 40 - Wis change was made to further clarify that the high reactor pressure, high drywell pressure, and reactor low water level and scram discharge level trip functions may be bypassed when in the " Refuel" mode with the reactor suberitical and the reactor coolant temperature less than 212'P as allowed by Note 3 of Table 3.1.1 (Page 31) . For clarity, NSP's proposed wording has been slightly modified.
5. Pages 43 and 45 - ne changes on these pages eliminates an unnecessary reference since the failure rate statistics are as listed in the Bnses immediately proceeding the reference.

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6. Page 59 - The changes in this page are a typographical correction and a change to allow bypass of the source range monitor (SRM) upscale block when the associated intermediate range monitor (IRM) range switches are above Position 6 instead of above Position 7.- De SRM Upscale rod block is required only when insufficient range overlap exists on the IRM hefore the SRM's reach the upper range l imit. At least one decade, but preferably two decades, of overlap is sufficient to assuro a definite overlap and transition from ,

monitoring of SRMs to monitoring IRMs. Since Position 6 corresponds \

to a three decade overlap, the change is considered acceptable. In i addition, operating experience has revealed that the original bypass k above the Position 7 is overly restrictive and interferes with normal *'

At ri et tin in ft manner that unu not I nt rin d eel .

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'f I 7. Page 62 - W e change on this page is a typographical correction.

1 j 8. Page 63 - We additional sentences are added to Note 2 (for Table 4.2.1) to define SRM and 1RM calibrations to avoid misunderstanding.

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9. Page 78 - An unnecessary date (May 1,1974) is removed from the f F"

limiting conditions of operation. D e surveillance requirements ,

are changed to require rod position verification prior to: insertion j i

as well as withdrawal of each rod group to correct ansmissida" error.

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10. Page 85A - De change on this page is a typographical correction.

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11. Page 39 - Le primer assembly contains two charges that tanno't be separated for test firing. We change reflects the reality of the single primer assembly containing two charges rather than independent charges as conceived when the Technical Specification was F written. We change eliminates the misleading impression that the explosive charges can or must be tested individually sad is

' acceptable.

' 12. Page 94 - Dilution from water in the cooldown circuit was included

- in the original determination of standby liquid control system requirements. W e change to the Technical Specification bases justifiably acknowledges this to avoid uncertainty about basic design information.

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13. Page 104 - Corrects numerical errors in converting absolute pressure j

to gaugt pressure.

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14. Page 108 -ne change on this page is a typographical correction.

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15. - Our evaluation of NSP's proposal for a surveillance l

Pages116-11{hangeisasfollows:

requirement c (1) Sampling of reactor coolant for radioiodines of I-131 through 1-135 instead of sampling for gross beta activity is approved based on the current Standard Technical Specification format.

5 (2) We instituted a limiting condition of 5 microcuries I-131 dose l

equivalent per gram of water to replace the 20 microcuries of total iodine per cubic centimeter of water. nis change was

'_, made to make the radiological consequences of a steam line break outside of containment acceptable and well within the limits of 10 CPR 100, i

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Additionally, we instituted surveillance requirements for steady state radioiodine concentrations prior to operation to be j

i consistent with that of sin 11ar boiling water reacter installa-  !'

tions, no change reflects a means for surveillance of possible ,

i transient radioiodine spike conditions during startup and 1

shutdowns.

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i Our other changes reduced the reactor coolant water chemistry conductivity and chloride limiting conditions in accordance with Reguistory Guide 1.56 (June 1973). The new limits reduce the j probability.of corrosion induced failures,in reactor coolant boundaries in boiling water reactors.- A reactor shutdown, if reactor coolant water purity limits are exceeded, assures curtail-

) ment of reactor operations until suitable corroetive measures '

are taken. All of the above mentioned changes were discussed i

with NSP (phone con of January 28,1975) and found to be acceptable b to NSP.

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16. Page 118 - The change improves clarity by adding "during power operation". NSP inadvertently left the item out of their original l

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submittal.

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17. Pages 132 and 133 - Our change provides the basis for the SUCi/ml f of water I-131 dose equivalent limit that replaced the 20pC1/m1 total iodine limitz *in Decification'3.6.C'1, i.e.; the steamline [

break accident outside of containment. It is in the standard i

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' Technical Specification format and has been reviewed and found acceptable by NSP.

18. Page 147A- D e change utilizes the Monticello Technical _ Specification '

l definition of "run mode" to simplify the specification operating

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i limit without compromising the basic requirements for containment l

l- i inerting as stated in 3.7.A.5.b. 1.: '

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19. Page 148 - The change permits filter tests 30 days prior to refueling so that related activities can be scheduled efficiently. The-existing requirement that filter performance tests be conducted during the i

' refueling outage is without basis and unintentionally restrictive..

20. Page 150 - No action was taken on this i.es because the typographical

' erm r did not appear on page 150 as revised with Change _No. 12 issued by the Cossaission on November 15. 1973.

21. his proposed change was disapproved for reasons stated in the introduction to this evaluation.

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22. Page 152 - The change in 3.7.D.2 to use "run mode" does not alter the intent of the limiting conditions of operation with respect to continued reactor operation with inoperable isolation valves.

The change in 3.7.D.3 corrects an oversight by properly referring to specification 3.7.D.1 and 3.7.D.2 instead of 3.7.D only.

23. Page 170 - The change on this page is a typographical correction in specification 3.8.A.9. The changes to specification 4.8.A.4 require accelerated surveillance at the point of increased iodine or particulate releases only, and are similar to surveillance require- ,

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ments for other BWRs with separate stack and vent releases.

24. Page 172 - The change eliminates specification 4.8.C.2e because ,

i.e., specification 4.8.C.1 is adequate. The j it is repetitious, NSP proposed page change mistakenly deleted 4.8.C.2.d although item 20 of Exhibit A of NSP's request dated November 15, 1974, correctly identified Specification 4.S.C.2.e. This has been confirmed by discussion with NSP representatives (phone con January 29, 1975).

25. Page 181 - The change restores the phrase "or the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" which was unintentionally deleted from the specification previously. The correction is required to specify the action that must be taken in the event that sources of electric power are unavailabic.
26. page 183 - The change requires a shutdown and cooldown to 212*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one 125V battery system or the 250V battery system is unavailabic. The proposed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown requirement is consistent with the safety importance of this system and justifiably more restrictive on plant operation than the 7 day period currently specified.

CONCLUSION We have concluded, based on the reasons discussed above, that because the changes do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the changes do not involve a significant hazards consideration. We also conclude that there is reasonable assurance (i) that the activities authorized by this amendment can be conducted without epdangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: APR 1: 0 N -

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