ML19319D304

Responds to 751103 Request.Forwards Reactor Sys Branch Comments on Tech Specs

ML19319D304
Person / Time
Site:	Crystal River
Issue date:	12/19/1975
From:	Baer R Office of Nuclear Reactor Regulation
To:	Mcgough J Office of Nuclear Reactor Regulation
References
NUDOCS 8003130955
Download: ML19319D304 (5)

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w Boseph M. McGough,' STS Group' Leader,. Operating Reactors, RL.

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THRU:' Thomas M. Novak ' Chief, Reactor Systems Branch, TR REACTOR SYSTITS BRANCR COM4ENTS ON CRYSTAL RIVER 3 TECHNICAL SPECIFICATIONS Per your request. (1tr. 4. McGough to Branch Chiefs dtd 11/3/75),

enclosed are the RSB comments on-the Proof 'and' Review Copy of the

- Crystal River 3 Technica1' Specifications.

C'M w W Robert L. Baer, Section A Leader Reactor Systems Branch Division ' f Technical Reviea.

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b RSB C0!t1ENTS ON CRYSTAL RI'/ER 3

' TECHNICAL SPECIFICATI0'Is 1.

LC0.3.5.1/4.5.1, page 3/4 5-1:

a) CFT Iqvel operating range shou.d also be expressed in terms of ft? of water to facilitate comparison to accident assump-tions.

Also, the lower of the allowed ocerating range should be the same value assumed for the LOCA analysis.

b) The lcwer of the allowed ooerating range for nitrocen cover pressure should be the sane value assumed for the LOCA analysis.

c)' The ' electrically-operated discharge valves from the CFT's shall be open and breakers locked onen and tagged.

d)

CFT appurtances which could allow the sourious opening of an electrically-onerated valve to comoronise CFT avail-i ability shall have such valves clnsed and breakers locked open and tagned (for examole, see vent line isolation valves HV 26511 and mi 26512).

- 2.

.LC0 3.5.2/4.5.2:

a) LTo be consistent with the recent recommendations made regarding the Standard Technical. Specifications, a single ECCS train is pernitted to be out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> instead of 48' hours.

b) A surveillance requirement to test the baron. dilution mode (s) of long term core ccoling should be implemented.

i c)

Page 3/4 5-5:

Paragraph d.

Add item'4 to read; i.

' Cycling each valve manually-(local handwheel) through at.least one complete cycle of full travel t'o verify the operability of the manual backup capability."

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d)

Page 3/4 5-7:

The lower of the allowed operating range of ppm boron in the BWST should be the same value assumed for

.the LOCA analysis.

The maximum value should not exceed the

. ppm assumed in. assessing the concentration levels during the-long-term after a LOCA.

e) Page B 3/4 5-2:

Paragraph 3/4/5/4.

Greater detail should be provided (or referencad) in this basis to show that 40 feet of water in the BUST is sufficient to allow adequate flPSH to the ECCS pumps when the shift to the recirculation mode occurs.

fl Why does 'the spec requi e MOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within toe next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, while the spec for an. inoperable CFT requires a HOT shutdown within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />?

3.

BASE _S_, page-8 3/4 1-1:

The relationship of the following sentences to the 15ak/k shutdown maroin requirement in specification 3.1.1.1 should be nade clearer:

"In the analysis of this accident a SHUTDOUM MARGli! of (0.4's)

.eik/k is initially required o control the reactivity transient.

Accordingly, the SHUT 00'.fi MARG!i! required is based upon this

. limiting condition and is consistent with FSAR safety analysis assumotions."

4. ' CASES, page B 3/4 1-1: This paragraph is not detailed enough to show where the required values of moderator tem 73rature coefficients in specification 3.1.1.3 were derived.

The vague. reference to'a " safety analysis" should be expanded 1to specify and discuss which analysis.

5.

LC0 3.1.2.5, page 3/4 1-12.

An coeratina DHR system is apparently necessary to ensure continuous mixing to allow the

-operator:a-prompt and definite indication (from the audible count rate instrumentation) of any boron dilution (uncontrolled or.otherwise).

In addition, an onerating DHR system is apparently assumed in LC0 3.9.1 which requires that-the boron concentration of all filled portions of the RCS-and the re-fueling canal be maintained uniform.

It therefore seems

-appropriate that for Mede 5 (Refueling), the DHR system n6t only be required to be OPERABLE, but also be required to be C.'ERATING.

This discussion should be in the SASES.

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Page 3/4 9-1.

Surveillance Requirement 4.9.1.2 allows a maximum time interval between samples of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

It should be con-firmed and put in the BASES that this test frequency is consistent with the time frame described for the dilution accident analyzed in the FSAR.

7.

Page 3/4 9-2.

Since the analyzed boren dilution accident would dictate that the operator have a prcmpt and definite indication of any boron dilution from the audible count rate instrumentation, it seems prudent to require that the more sensitive channels temporarily installed during initial core loading provide audible indication in the control room.

It is not obvious that sufficient sensitivity of the excore source range instrunentation exists during initial core loading to promptly alert the operator of a continuing boron dilution.

Also, since the availability of such instrunentation is of safety significance, we should also require redundancy in the control rocm.

8.

Page 1-3, definition 1.14b.

The word "both" is confusing.

Should read as follows:

b.

Leakage into the containrent at.csnhece '<.hich is kr.com not to interfere with the operation of leak detection systens and which cccurs at known locations exclusive of PRE 35URE ECUNDARY Lt A:%3E...

9.' Page 1-3, cafinition 1.15.

The de'inition of L"! IDENTIFIED LEIK/ 3E as given inc;uces intersystea leskage, lich is not co-sistent with Regulatory Guide 1.45.

Recomend a separate definition of Intersystem Leakage for paragraph 1.15; rove U'l;DENTIFIED LEAKAGE to paragraph 1.16.

INTERSYSTEM LEAKAGE 1.15 INTERSYSTEti LEAKAGE shall be all leakage frcn the Reactor Coolant Pressure Boundary to seccndary coolant or auxiliary systens across passive barriers such as valves or heat exchangers.

UNIDENTIFIED LEAKAGE 1.16 UNICENTIFIED LEAKAGE shall be all leckage which is not IDENTIFIED LEAKAGE, CONTROLLED LEAKAGE, OR INTER-SYSTEM LEAKAGE.

(It is reccmmanded that the existing definitions 1.16 and 1.17 precede existing definition 1.14).

74

'10. 'Page 3/4 4-9, paragraph 4.4.6.2 - The surveillance require-

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-ments of this paracraoh seem inconsistent with the Recclatory Guide ~1'.45 criterion of detection of cne gpm in one hour or less.

If.the monitoring systems are provided with alarms

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as also recommended by R.G. 1.45, cerhaps all of the in-dic'ated surveillance actions should also be required on

receipt of an alarn.

Ih Limiting Conditions for Operation should also include the

'following:

_(1) Total Core Power - Licensed core power level by heat balance determination (which assumes 2*; measurecent un-certainty);.coatinuous monitoring instrumentation and the frecuency of.recalibration to the heat balance should be specified.

~(2)

Reactor Vessel Cociant Mass Flow Rate - The minimum flow initial condition assured in the 0:DR safety evaluation for each permissable 1000 operatino conficuration should be snecified as a minimum value.

The specified flow should include a uncertainty allowance based on identified

-monitoring instrumentation.

The taximum flow rcte should also be specified.

The value selected should be justified in the CASES, con:idering the :otentici for core lif t cnd any other safety considerations resulting from high reactor flew rates.

' (3)' React,o_r Core Ccolant AveraSe Tencerature - The icwest initial valve assumed for LOCA and/or anticipated tran',ient analyses -starting at full power should be specified as a maxiru-1 value.

Monitoring instrumentation and the corresponding uncertainty allowance should be specified.

(4)

_ Pressure in the Pressurize _r_ - The nornal oressure control i

band shoulc be soccified since this was the basis for initial conditions ~ assumed for the safety evaluation.

Monitoring instrumentation and associated uncertainty should be-specified.

'12.x BASES 3/4.1.3 'and 3/4.2.

These bases should more clearly relate'

.to the LOCA analysis and should include a plot of allowable LHGR versus Core Height as a major aspect of discussion.

l 13.. LCO 3/4.1.1.

A LOCA analysis has been subcaitted for partial loop operation with 3 reactor coolant pumps.

L'ntil an analysis is submitted for 2-pump operation, Technical Specification must prohibit this.'co_nfiguration.

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