ML19256E778
| ML19256E778 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/12/1979 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML19256E775 | List: |
| References | |
| NUDOCS 7911150196 | |
| Download: ML19256E778 (13) | |
Text
{{#Wiki_filter:.. .. EXHIBIT A Prairie Island Nuclear anerating Plant License Amendment Request dated November 12, 1979 Proposed Changes to the Technical Specifications Appendix A of Operating Licenses DPR-42 and -60 Pursuant to 10 CFR 50.59, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications.
- 1. Maximum Coolant Activity PROPOSED CHANGES Revise Section 3.1.0 as shown on attached Exhibit B pages TS.3.1-11.
12, - 13, and Figure TS 3.1-5. REASON FOR CHANGE This change is in accordance with a request of the NRC staff to modify RCS activity specifications to comply with the standardized technical specifications. This specification incorporates the definition of DOSE EQUIV/LLENT I-131 and revised definition of E. The action statements on pages 3/4 4-22 and -23, Figure 3.4-1, and basis on pages B 3/4 4-5 and
-6 of the standardized technical specifications have been incorporated directly with minor wording changes made to clarify intent and to put the specification in the same format as the existing Prairie Island technical specificatons. The threat of release at cold shutdown is minimal thus 4(a)'s sampling requirements (Table 4.4-4 and Section 3.4.8 Action Statement) were changed accordingly (see Section 3).
SAFETY ANALYSIS This specification imposes new requirements and action statements con-sistent with NRC regulatory guidance. These specifications are based on ensuring that the resulting 2 hour dose at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident with the activity limits specified. 1336'213 7911150
/96%
- 2. Steam Generator Activity Limit (Section 3.4)
PROPOSED CHANGE On page TS.3.4-2, change specifications 3.4. A.9 and 3.4.B.1 to read as shown on the Exhibit B page. On page TS.3.4-3, change the basis for the steam generator activity specification to that shown on the Exhibit B page. REASON FOR CHANGE This change was requested by the NRC staf f in an October 12, 1979 meeting with NSP personnel. SAFETY ANALYSES The existing Technical Specifications limit steam generator I activity to 0.3 uc/ml. If a steam line break occurs such that the contents of one steam generator is released to the environment, the predicted site boundary dose would be 1.5 rem, consic rably below the 10 CFR 100 limit. In this revision the limit would be 0.1 uc/gm dggg equivalent I-131 which corresponds to approximately .06 ue/ml I . Thus the limit is effectively being reduced by a factor of 5 (a more con servative direct ion) . The standardized Technical Specification limit is site independent, but on the basis of the foregoing discussion, it is adequately conservative.
- 3. Chemistry / Radiochemistry Sampling Requirements (Table TS.4.1-2B)
PEOPOSED CHANGE In Table TS.4.1-2B, change the reactor coolant and secondary coolant samp1-ing requirements to that shown on the attached Exhibit B Table TS.4.1-2B pages. On the Exhibit B test items 1, 5-13 are unchanged from the previous page revision, except to condense the title of the test from ? columns to one and to use the terms "RCS" fer reactor coolant, "Rt3T" for refueling water storage tank, " Caustic Ste.ndoipe" for Chemic al Additive Tank. Items 2-4, 14 and 15 are new or changed requiremt r:s. These items have been incorporated directly from the standardized echnical specification with the following exceptions-1336 214
Test # Comment
- 1. The sampling frequency has been lef t at 5/ week which is more frequent than required by the STS and is consistent with current manning requirements.
4a The RCS isotopic analysen are to be conducted above cold shutdown . The STS requirements appear to require analyses at or ebove cold shutdown. Since the unit could be taken to cold shutdown as a followup corrective action, continued monitoring on a 4 hour basis serves no usc ful purpose. Rather the emphasis should be to identify the cause of the high activity and either correct the cause or implement methods to reduce the activity. Sampling would be as appropriate rather than on the arbitrary basis specified. 4b. The requirement for sampling af ter a 15% load change has been changed to 25% for the following reasons:
- 1. Iodine sampling conducted during selected load follow operations over the past 3 years has indicated no particular sensitivity to 15% load changes compared to 25% load changes.
- 2. Data taken by the plant staf f during load follow operations and after reactor trips has been used to characterize peak / base ratios and shown that the major parameter af fecting the peak height is % fuel defects (for load follow operations only).
- 3. The Prairie Island units frequently load follow >15%/hr on back shifts. This would require personnel call out for analyses. Load changes of >25%/hr are less frequent or can be avoided under most circumstances.
REASON FOR CHANGE These changes were requested by the NRC staf f in an October 12, 1979 meeting with NSP personnel. SAFETY ANALYSES The new surveillance requirements will provide added assurance that regulatory limits will not be exceeded in the unlikely event of a steam generator tube rupture or steam line break as appropriate. 1336 215
EXHIBIT B License Amendment Request dated November 12, 1979 Do:ket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exhibit B conaists of revised pages of the Prairie Island Nuclear Generating Plant Technical Specifica-tions, Appendix A, as listed below: Pages (TS-) iv 3.1-11 3.1-12 3 1-13 3.4-2 3.4-3 Table TS.4.1-2B (2 pages) and adds a new figure: TS.3.1-3
}3)b'-
TS-iv REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3 1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RT for Reactor Vessel Steels Exposed to 550 Temperature DT 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3 10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope For F = 2.21 3.10-6 Deviation from Target Flux Difference as a Function of 0Thernal Power 3.10-7 Rod Bow Penalty (RBP) Fraction Versus Kagion Average Burnup 3.10-8 V(Z) as a function of core height 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Ilant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-site Operating Orgsnization 6 1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Opsrating Group l) O'2l7
TS.3.1-li REV D. HAXIMUM COOLANT ACTIVITY Specification:
- 1. The specific activity of the primary coolant shall be limited to:
(a) Less than or aqual to 1.0 microcuries per gram DOSE EQUIVALENT I-131, and (b) Less than or equal to 100/E microcuries per gran.
- 2. In Specification 3.1.D.1 the following definitions apply:
(a) DOSE EQUIVALENT I-131 is that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-lii, and I-135 actually present. The e' 'id dose conversion f actors used for this calculation s' chose listed in Table III of TID-14844,
" Calculation of L stance Factors for Power and Test Reactor Sites."
(b) E shall be the average (weighted in proportion to the concentra-tion of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamna energies per disintegration (in MeV) for isotopes, other than iodines, with half lives grear.er than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
- 3. If a reactor is above hot shutdown and RCS reaperature is greater than or equal to 500 F:
(a) With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure TS.3.1-5, operation may continue for up to 48 hours provided that the cumulative operating time under these cir-eumstances does not exceed 800 hours in any consecutive 12-month period. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-month period, a special report to the Commission shall be submitted within 30 days indicating the number of hours above this limit. (b) With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure TS.3.1-5, the affected reactor shall be shutdown and RCS temperature cooled to 500 F or less within 6 hours. (c) With the specific activity of the primary coolant greater than 100/E microcurie per gram, the affected reactor shall be shutdown and RCS temperature cooled to 500 F or less within 6 hours of detection. 1336'218
TS.3 1-12 REV
- 4. If a reactor is above cold shutdown:
(a) With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E_ microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits. A reportable occurr:nce re-port shall submitted to the Commission within 30 days. This report shall contain the results of the specific activity analyses together with the following information:
- 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded,
- 2. Fuel burnup by core region,
- 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded,
- 4. History of de-gassing operations, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
- 5. The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131.
_ Basis The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Prairie Island site, such as site boundary location and meteorological conditions, were not considered in this evaluation. Specification 3.1.D-2, permitting power operation to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure TS.3 1-5, accommodates possible iodine spiking phenomenon which may occur following changes in thermal power. Operation with specific activity levels exceed-ing 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3 1-5 must be restricted to no more than 800 hours per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by
, k f) b'-
TS.3.1-13 REV Figure TS.3.1-5 increast the 2 hours thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. The reporting of cumulative operating time over 500 hours in any 6 month consecutive period with greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131 will allow suf ficient time for Commission evaluation of the circumstances prior to reaching the 800 hour limit. Reducing RCS temperature to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements in Table TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isctopic analyses following power changes may be permissible if justified by the data obtained.
, )f) b-
Figure TS.3.1-5 REV
;.a .. 7 . .,.j. {, j g.j. -y . 4. J j E 4_,..._,
;J . J
}
.l1!.._a. j...
4_ 7_. g.
.d a . ,. p k,,.;.j .
,J.. _{_;, ..
i. _l_, 3 7 0 250 ' l_ _'
,3 l .. . , . 4_ _4 .,! .l j _ . .
[_ a_ _ _ _ _ 2_ _.. ,_ 7 _ j .}.. .p j :_1 1 4.;L q._ _ _ L q_ 4 lg . . .! _4 _, . t . _ _ .9 .__.._ . _
- 9. . .p
] . j _1.. .
.p{. }_ .
..y .
,.y. ..$. _.
.s_._ t.. _ . . _._ .
..q.,_
7.. _p. _. . _ . _ _ _ _ _ _ . . _ __ _ _
- 3. . g . 7
_ .,_._ w_ _ 9 7_. 9 _ .. ._ . _. . . _ . .. . _ .__
> ., . , _ p_ .
_ l .. I . a , ; _ . __ _. _ h200 , ,; ;. , ; ,, _L
,4,; {_
UNACCEPTABLE ._ .. 4 q.t_p, ..q j . , .. . _j. l . . . ..._ _. - g ..,_,9 .,_, . q ._
- 7. OPE. RAT, ION . . _ _
_. .; . . ..t .;{l ....
., I .. ..
_T .
. _ __ 4 . .. _
. [_ . _ _ _ _ _
S _t43. . t_1 4._
.L j .]..,,.{-i ,--
1 o 4.. - - -- . .
-t ,
]._..
.j 9.. . j j. . _ F_ . j . _. . . _-
-{ _ _1. ... _ .
.3,
.j.
3 _.. . . . . j.;.. _ .
.,. 3 .,. _ _ _
z J_ q . q _ j . .. -+pI.;j . ; ;_ . .. __ _ . _ u . _ .. _ __ _ _
$ , . . 7_ .n--t- t- - '~ - - -
-lI, - - - --- -
O 7.s,. _ .7 _1j.. o .. .' ,g.t.
.i._,. , t '. .-_ _- _.1. . !. _
_.,.J. o _. ,_ _
..p j..j_ 4 _q __. _ _.._
J. _ _
} l 1.1 _
_ ,_ 9 _ p , .. J . _..;l _._; _ _p.H .. _
.l_. . . ___
.] _ _ _ __ . . . .J. . . . .
g 7, p.. , _ _-4_3 p. _ .. .. _ _q .
..l _ 7_
.-; _ _a 3_ _9_
.p. . . _ _
g . .
,. 4 _.
9
- 100 g .} .; _ } _ ..{__ _ . !_
_{ g;. Jj.: _p . _& j . _ F_ . ., . _p _
.g_ ;_ 7 g
_.l__..
. p. ..
.q..
.p g
.l.._
g..,r. - u. q _ _ . . i pl_M, f~ a_ _f %_. _ _.Pr -l-.-
, q .; . __ _ . ._ . _ _ _
$ 1..E'Jf- ACCEPTABLE _ Z
-~~
--\ - - "
~' T d- -
uj 50 I H-f OPERATION - - q-- - - - - - - - - J-f '- - - 4 f _j_. q
) _ f. . a._4 L . .j _l .
_T_._{,_._ m c
. e j .}. __._ 7._.
7 , q _._
, __ __3 _.
~~
_4_ _H__ y -*- ---
- . 7_, . .
.p. .
.T.; . _... _
'_T . 7 _ ._ __ . . __.
uJ ; - -. , - - - - g- - - - - -
+- -- 7, - - - - - --
_- _ . - .- ..j . ' 1_ _ L ) . ' ., . to 9_L 4 q_ .,. ~. - .. 4- _g_ . _.
.1 i 8 t 7__,_ .;it! - i-I--
- t t--
_ L , .. _1,- -. . _ f, . .}_ _- _-_._j_. 7 . ,_
. r 1- ,i t 0
20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER 1 336'2,21 FIGURE 3.1-5 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0gCi/ gram Dose Equivalent 1131
TS.3.4-2 REV
- 8. Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line rupture outside of containment shall be operable or at least one damper in each duct shall be closed.
- 9. The specific activity of the secondary coolant system for that reactor shall be < 0.10 uCi/gm DOSE EQUIVALENT I-131.
l B. If, during startup operation or power operation, any of the conditions of Specification 3.4 A., except 2.b.or 9., cannot be met startup l operations shall be discontinued and if operability cannot be restored withh 48 hours, the affected reactor shall be placed in the cold shutdown condition using normal operating procedures. If 2.b. is not met within 7 days, one unit shall be placed in the cold shutdown condition. If 9. is noc met, the affected reactor shall be placed in hot standby within 6 hours and cold shutdown within the following 30 hours. Basis A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the con-denser and by continued feedwater fiov Lo the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system. The ten main steam safety valves have a total combined rated capability of 7,745,000 lbs/hr. The tota'. full power steam flow is 7,094,000 lbs/hr; th ere fore , the ten main steam total steam flow if necessary.gety valves will be able to relieve the In the unlikely event of complete loss of electrical power to either or both reactors, continued removal of decay heat would be assured by avail-ability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the main steam safety valves. One auxiliary feedwater pump can supply suiticient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedwater pump for each reactor can be made available to the other rcactor. 1336 222
TS.3.4-3 REV The minimum amount of water specified for the condensate storage tanks 's sufficient to remove the decay heat generated by one reactor in the fir st 24 hours of snutdown. Essentially unlimited replenishment of the cona.nsate storage supply is availble from the intake structures through the cooling water system. The two power-operated relief valves located upstream of the main steam iso.1.ation valves are required to remove decay heat and cool cne reactor down following a high energy line rupture outside containment. Isole lon dampers are required in ventilation ducts that penetrate those rooms containing equip-ment needed for the accident. The limitations on secondary system specific activity ensure that the re-sultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary - 5e leak in the steam generator of the af fected steam line. These values srs con-sistent with the assumptions used in the accident analyses. Reference (1) FSAR, Sec. ion 10.4 13315'223
Table TS.4.1-2B Page 1 of 2 REV TAB' E TS.4.1-2B MINIMUM FREQUEPCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Reference
- 1. RCS Gross 5/ week Activity Determination
- 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVALENT I-131 Concentration
- 3. RCS Radiochemistry EI detarmination 1/6 months (l) (when at power)
- 4. RCS Isotopic Analysis for Iodine a) Once per 4 hours, wheneve r Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE EQUIVALENT I-131 or 100/E uCi/ gram (above cold shutdown), and b) One sample between 2 and 6 hours following a THERMAL POWER change exceeding 25 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown)
- 5. RCS Radiochemistry (2) Monthly
- 6. RCS Tritium Activity Weekly
- 7. RCS Chemistry (C1, F, 0 )* 5/ Week 2
- 8. RCS Boron Concentration *(3) 2/ Week (4) 9.2
- 9. RWST Boron Concentration Weekly
- 10. Borie Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentration Monthly 6.4
- 12. Accumulator Boron Concentration Monthly 6
- 13. Spent Fuel Pit Boron Concentration Monthly 9.5.5 1336'??4
.. v. Table TS.4.1-2B Page 2 of 2 REV l TABLE TS.4.1-2B MINIMUM FREQUENCIES FOP SAMPLING TESTS FSAR Section TEST FREQUENCY Reference
- 14. Secondary Coolant Gross Beta- Weekly Gamma activity
- 15. Secondary Coolant Isotopic 1/6 months (5)
Analysis for DOSE EQUIVALENT I-131 concentration NOTES:
- 1. Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was last suberitical for 48 hours or longer.
- 2. To determine activity of corrosion products having a half-life greater than 30 minutes.
- 3. See Specification 3.8 for requirements during refueling.
- 4. The maxinum interval between analyses shall not exceed 5 days.
- 5. If activity of the samples is greater than 10% of the limit in Specification 3.4.A.9, the frequency shall be once per month.
- See Specification 4.1.D.
1336'225}}